NSD-NRC-98-5543, Forwards Westinghouse Responses to FSER Open Items on AP600. Summary of Encl Responses Provided in Table 1
| ML20199F355 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 01/28/1998 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD-NRC-98-5543, NUDOCS 9802030149 | |
| Download: ML20199F355 (51) | |
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Westinghouse Energy Systems b 355 Electric Corporation " "5 *'8 ' " 8 9 '
- 15230 0355 ,
- DCP/NRCl232 NSD-NRC 98 $$43 Docket No.
- $2 003 ,
! < January 28,1998 3
Document Control Desk i U.S. Nuclear Regulatory Commission '
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Washington, DC 20$$$
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' A'ITENTION: T. R. QUAY SUlijECT: AP600 RESPONSE TO FSER OPEN ITEhiS
Dear hir. Quay:
Enclosure 1 of this letter provides the Westinghouse responses to FSER open items on the AP600. A summary of the enclosed responses is provided in Table 1, included in the table is the FSER open item number, the associated OITS number, and the status to be designated in the Westinghouse status column of OITS.
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The NRC should review the enclosures and inform Westinghouse of the status to be designated in the
! "NRC Status" column of OITS.
! Please contact me on (412) 374 4334 if you have any questiens concerning this transtalttal.
l.
Brian A. hiclntyre, innager h6/A f'
. Advanced Plant Safety and Licensing i
j jml Enclosure l cc: W. C. Iluffman, NRC (Enclosure)
- T. J. Kenyon, NRC (Enclosure)
! J. hi Sebrosky, NRC (Enclostire)
D. C. Scaletti, NRC (Enclosure)
N. J. Liparuto, Westinghouse (w/o Enclosure) 9902030149 990128 -
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DCP/NRCl232 NSD NRC 98 5543 2- January 28,1998 Tahic 1 List of FSER Open items included in latter DCP/NRCl232 FSER Open item OITF Number Westinghouse status in OITS 220 '25F (R2) 6312 Confirm W 230.146F 6505 Confirm W 280.32F 6498 Confirm W 410.339F (RI) 6194 Confirm W 480.873 (R1) 4915 Action N 650.llF(RI) 5973 Action N i 720.440F (RI) 6178 Confirm W 720.441F (RI) 6179 Action N
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Enclosure to Westinghouse Letter DCP/NRCl232 January 28,1998 l
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FSER Open P'm ..,
i j i Open item 220.125F (OITS #6312) Response Revision 2 Because a massive amount of water in to be contained in the PCCWS tank, the staff raised a concern that the COL applicant should monitor the venical and radial deformation of the tank during initial filling, and compare the measured values with the tank deformation predicted by calculation. He staff identified this 1, sue as Open item 3.8.4.4 3 and COL Action item 3.8.4.41.
At the meeting on June 12 through 16,1995, Westinghouse stated that the water weight is sma'), in comparisor with the total weight of the shield building roof structure (estimated to be about 10 percent). Westinghouse also showed that the denection of the roof stmeture resulting from the first fill of water should be negligible. On that basis, Westinghouse contended that there is no need to monitor the tank denections and compare the denections against predictions.
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During the meeting on December 9 through 13,1996, Westinghouse repeated its justifie.' ion concerning this issue. Ilowever, the staff did not agree with Westinghouse % basis for not monitoring the venical and radial deformation of the tark during initial tank filling. Moreover, the staff assened that post construction testing is necessary to confirm the adequacy of the PCCWS tank. This is because the st ff's review experience suggest that the excessive deformation resulting from the massive amount of water may cause cracking of the tank wall and base slab, as well as water leakage from reinforced concrete tanks with steel liners, in Revision 17 of SSAR Section 3.8.4.1.1, Westinghouse added a statement that leak chase channels are provided over the liner welds to permit monitoring for leakage and to prevent degradation of the reinforced concrete wall which might result from the freezing and thawing of leakage. Also.
Westinghouse indicated that the exterior face of the reinforced cc7 crete boundary of the PCCWS tank is designed to control cracking, in accordance with Paragraph 10.6.4 of ACI 349, with reinforcement steel stress based on sustained loads (including thermal effects). However, Westinghouse still did not commit to monitor the vertical and radial deformation of the tank during initial filling and compare the n,wasured values with the tank deformation predicted by analysis. On the basis of the above discussion, the staff concluded that Westinghouse's response to the staff': concem (as stated in Revision 17 of SSAR Section 3.8.4.1.1)is not acceptable, nerefore, Open item 3.8.4.4 3 and COL Action item 3.8.4A 1 remair, unsolved.
I Response (Revision 2):
De SSAR is revised below to show monitoring of the tank during initial filling. Requirements for visual smination are given. The calculated deflections of the roof structure due to the Orst fill of water s: than one quarter of an inch. Monitoring of tank denections and comparison against prediet is difficult because of the small magnitude of the denections due to the water inventory.
Venicas denections could also be caused by thermal changes. The venical denection will be measured during tank till and will be compared to the predicted magnitude. His will be used in combination with the visual examination to confirm acceptability.
SSAR Revision:
SSAR changes included in Response Revision 2 include a sentence added to address monitoring of structures during operation. His was added in response to discussion with NRC staff.
M@N 220.125(R2) 1
. FSER Open 1::m ,, m
!!! i:3 Revise subsection 3.8.4.7 as follows:
3.8.4.7 Testing and In-Service Inspection Requirements x I Structures supporting the passive containment cooling water storage tank on the shield I building roof will be examined before and after first filling of the tank.
I I
- The boundaries of the passive containment cooling water storage tanic and the tension ring of the shield building roof will be inspected visually for any -ign, cf !cakege-er i distress excessive concrete cracking before and after first filling of the tank. Any I significant concrete cracking will be documented and evaluated in accordance with ACI 349.3R 96 (reference 50).
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- Re vertical elevr. tion of the passive containment cooling whier storage tank relative to I the top of the shield building cylindrical wall at the tension ring will be measured I before and after first filling. The change in relative elevation will be compared against i the predicted deilection.
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- A report will be prepared summarizing the test and evaluating the results.
I There are no other in-service testing or inspection requirements for the seismic Category I -
I structures. However, during the operation of the plant the condition of these structures /
I should be monitored by the Combined License applicant to provide reasonable confidence I that the structures are capable of fulfilling their intended functions.
- Revise subsection 3.8.6 as follows. This includes revision shown in response to Open item 220.119.
3.8.6 Combined License Information This-=ction har. nc requirernent for additica:! informatica te be provided in-support of the Cc=hined Licenz application 1 3.8.6.1 CorNinment Vessel Design Adjacent to Large Penetrations l
I The final design of containment vessel elements (reinforcement) adjacent to concentrated I masr* 1 penetrations) is completed by the Combined License applicant and documented in I the A.iME Code design report.
I 3.8.6.2 Passive Containment Cooling System Water Storage Tank Examination i ne Combined' Licente applicant should examine the structures supporting the passi e I cor.tainment cooling storage tank on the shield building roof during initial tank filling as i described in subsection 3.8.4.7.
Add to reference i Subsection 3.8.7 the following:
I 50 ACI 349.3R 96. " Evaluation of Existing Nuclear Safety-Related Concrete Structures"
[ W8Silligh0USe 220.125(R2)-2
,._ FSER Open item ,
230.146F (OITS #6505) 10 CFR Part 50, Appendix S, Section III, requires that the integrity of the reactor coolant pressure boundary be maintained following a safe shutdown earthquake (SSE). He staff is concemed that the CVS piping is non seismic and, consequently, its integrity cannot be assured following an SSF. The staff believes that the CVS piping should be designed to withstand an SSE without failure.
Westinghouse should commit 'that the CVS system within containment will be designed, analyzed, and constructed to meet the seismic requirements of Seismic Category Il structures, systems, and components (SSCs). The staff recognizes that ordinarily, for the CVS to be categorized as Seismic Category II, failure of the CVS SSCs would have to compromise the function of some AP600 safety related SSC. Houver, to assure the seismic integrity of the CVS, Westinghouse should commit to designing, analyzing, and constructing the po tion of the CVS within containment to the same criteria as that used for Seismic Category 11 SSCs, ami document this in both the SSAR and the Certified Design Material.-
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Response
To address the concems of the staff, a seismic evaluation will be performed for the non seismic portion of the CVS system inside containment to provide for the seismic integrity and pressure boundary integrity of the corresponding piping designated as reactor coolant pressure boundary.
Section 5.2.1.1 of the SSAR will be resised as indicated below to identify the specific criteria to be used for this evaluation. His criteria is based on Equation 9 of the ASME Code,Section III, Class 3.
Fabrication, examination, inspection, and testing req irements as defined in Chapters IV, V, VI and VII of the ASME B31.1 Code are applicable and will be used for the corresponding B31.1 (Piping Class D) CVS piping systems, valves, and equipment.
Section 2.3.2 of the AP600 Cenified Design Material will be revised as indicated below to confirm that the seismic analysis has been performed and structural integrity maintained for the corresponding non-safety related CVS piping.
SSAR Revision:
Add the following to subsection 5.2.1.1:
I Seismic Integrity of the CVS System Inside Containment I
i To provide for the seismic integrity and pressure boundary integrity of the non-safety I related (B31.1, Piping Class D) CVS piping located inside containment and designated as I reactor coolant pressure boundary, a seismic analysis will be performed and a CVS I Seismic Analysis Report prepared with a faulted stress limit equal to the smaller of 4.5 Sh
.I and 3.0 Sy and based on the following additional criteria :
I I Additional loading combinations and stress limits for nonsafety-related chemical and i volume control system piping systerr.s and components inside containment
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.. FSER Open ihm m
i Equations I Condition Loading Combination (3) (ND3650) Stress Limit i
I Level D PMAX(I) & DW + SSE + SSES 9 Smaller of I 4.5 S h r 3.0 S y i
I SSES FAM/AM 1.0 S h l
I TNU + SSES i ( M1 + M2)/Z(2) 3.0 S h l
I Notes:
1 1. For earthquake loading, PMAX is equal to normal operating pressure at 100% power.
I I 2. Where: M1 is range of moments for TNU, M2 is one half the range of SSES I moments, 1 M1 + M2 is larger of Mi plus one half the range of SSES, or full range of i SSES.
I 1 3. Sr T21e 3.9 3 for description of loads.
I 1 4. FAM is amplitude of axial force for SSES; A g si nominal pipe metal area.
I I Component supports, equipment, and structural steel frame are evaluated to demonstrate I that they do not fail under seismic loads. Design methods and stress criteria are the same I as for corresponding Seismic Category I components. The functionality of the chemical I and volume control system does not have to be maintained to insuit structural integrity of I the components.
I i Fabrication, examination, inspection, and testing requirements as defined in Chapters IV, i V, VI, and VII of the ASME B31.1 Code are applicable and used for the B31.1 (Piping i Class D) CVS piping systems, valves, and equipment inside containment.
ITAAC Revision:
Section 2.3.2 Chemical and Volume Control System Insert Item 15.
- 15. The non safety relt.ted (B31.1, Piping Class D) piping located inside containment and designated as reactor coolant pressure boundary, as identified in Table 2.3.2-2, has been designed to withstand a seismic design basis event and mamtain structural integrity.
[ W85tlDgh0Use 230.146F-2 l
. FSER Open item y
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11AAC Revision:
Section 2.3.2 Chemical and Volume Control System insen item 15.
- 15. The non-safety related (B31.1, Piping Class D) piping located inside containment and designated as reactor coolant pressure toundary, as identified in Table 2.3.2 2, has been designed to withstand a seismic design basis event and maintain stmetural integrity.
Table 2.3.2-2 Insert the following Line Numbers Table 2.3.2-2
- Line Name Line Number ASME Code Section 111 CVS Supply Line to Regenerative BBD LOO 2 No '
IIcat Exchanger CVS Retum Line from Regenerative BBD L018 No lleat Exchanger BBD LO73 No CVS Line from Regenerative Heat BBD LOO 3 No Exchanger to Letdown Heat BBD LO72 No Exchanger-CVS Lines from Letdown Heat BBD LOO 4 No Exchanger to Demin. Tanks BBD LOO 5 No CVS Lines from Demin Tanks to RC BBD LO20 No ,
Fillcrs BBD LO2l No BBD LO22 No BBD LO29 No BBD LO37 No CVS Lines from"RC Filters to BBD LO30 No Regenerative lleat Exchanger BBD LO31 No BBD LO34 No No CVS Resin Fill lines to Demin. Tanks BBD LOO 8 No BBD LOl3 No BBD LO25 No D# 230.146F-3 1
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., FSER Open lum R,,
Design Commitment Inspections, Tests, Analyses Acceptance Criteria
- 15. The non safety related Inspection will be conducted of The CVS Seismic Analysis i (B31.1, Class D) piping the as-built components as Reports exist for the s located inside containment and documented in the CVS non safety related designated as reactor coolant Seismic Analysis Report. (B31.1, Class D) piping pressure boundary, as located inside containment and identified in Table 2.3.2 2, has designated as reactor coolant been designed to withstand a pressure boundary as seismic design basis event and identified in Table 2.3.2 2.
maintain structural integrity.
W W Westinghouse 230.146F-4
NRC FSER OPEN ITEM Open item 280.32F (OITS #6498)
Fire Protection of Safa Shutdown and Cold Shutdown Capability The NRC staff has developed the foll0 wing criteria for the protection of safo and cold shutdown capability following a singte fire in any fire area of the AP600:
Safe shutdown following a fire is defined for the AP600 as the ability to achieve and maintain the RCS temperature below 420 of without venting of the primary coolant ficm the reaciar coolant system (RCS). This is ;i departure from the enteria applied to te evolutionary plant desirjas, and the existing plants where sala shutdown for fires applies to both ho and cold shutdown capability. Cold shtttdown for the AP600 is defined as the ability to achieve and maintair the RCS below 200 oF, consistent with the enteria applicable to the evolutionary designs and existing plants. The use of the non safety related -
normal shutdown systeras and/or the safety related passive systems are acceptable to the staff to achieve and maintain safe shutdown following a fire. The safety related passive systems are considered an attemate shutdown method as described in Branch Technical Position (BTP) CMEB G.51. Consistent with the fire protection criteria for the advanced light water reactors specified in SECY 90-16 and SECY 93 87, redundant divisions of these systems shail be separated such that a fire in any fire area outsids of the containment or the main control room will not impair th9 plant's capability to achieve and maintain safe shutdown as deficed above, assuming a loss of all equipment in the ahected fire area. Consideration in the safe shutdown analysis upon personnel entry into the affected fire area to ropair or operate equipment to achieve safe shutdown is prohibited as presenbed in SECY 90-16. Personnel entry into the affccted fire area to repair or operate equipment necessary to achieve and maintain cold shutdown of the AP600 is acceptable, due to the unique capability of the APG00 to remain in safe shutdown using only passive systems for an extended period of time. The entena concerning cold shutdown capability ceviates fram the e deria applied to the evolutionary reactor designs, but are consistent with the criteria applicable to existing plants. To enhance the survivability of the normal safe shutdown and cold shutoown capability in the event of a fire, and to reduce the reliance on the infrequently utilized safety related passive systems, automatic suppression snall be provided in those fire areas outside containment where a fire could damage the normal shutdown capability, or result in a spu6ous operation of equipment that could result in a venting of the RCS. This criterica is unique to the AP600 and does not ensure that the normal shutdown capability will be free of fire damage or that the equipment necessary to achieve and maintain cold shutdown can be repaired within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Due to the inability of the fire brigade to rapidly enter the AP600 containment in the event of s fire, and the potential for damage to safety related and no mal shutdown equipment, in addition to potential rpuncus actuation (s; resulting in a venting of prirnary coo! ant from the RCS, the protection of circuits a.1d equipment insida c]ntainm3nt should be enhanced bevond the criteria specified in BTP CMEB 9.5.1 for existing plants consistent with the staff's technical position stated in Section 0.3 of NUREG-1242, "NRC Review of Electric Power Research Instituto's Advanced Light Water Reactor Utihty Requirements Document, Volume 3. Passive Plant Designs," published August 1994.
Several areas outside containment cordaining equipment and cables necessary to achieve and maintain safe and cold shutdow.) using the normal systems have not been provided with automatic suppression in accordance with the above enteria. Systems listed in Section 7.4.1.3 of the SSAR as recessary for 280.32m
.w wesuneme
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NRC FSER OPEN ITEM E "tig l i normal shutdown that have not been provided with automatic suppression include the chemical and r
volume control system, chilled water system, control power, compressed air, instrumentation, and steam >
Generator power operated relief valves.
Section 9A.2.7.1 of tne SSAR sicles that manual throttling and closing of the first stage automatic depressurization (ADS) valve to reduce the RCS pressure to the cperatiiig pressure of the normal residual heat removal system (RNS) will be used. The use of the first stage ADS results in a venting of the RCS to the in conta'nment refueling water storage tank (IRWST) in conflict with the fire prote, tion criteria established for safe chutdown of the AP600. In a letter cated October 14,1997, the applicant submitted its description of the AP600 response to a loss of oitsRe power with ' realistic" assumptions using the LOFTRAN code. The !oss of offsite power (LOOP) arw,ysis is used as a surrogate for the plant response following a fira as it can cause similar failure modes as the LOOP. In the analysis the applicant states that after approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the RCS hot leg temperature is less than 420 oF and the pressurizer pressure is approximately 600 Pia. A RELAPS analysis prepared by the Anatytical Support Group, SASG 95-02, dated May 1995, indicates that after approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> the RCS hot leg temperature will be approximately 430 oF and the pressunzer pressure is approximately 1600 psig.
Both of these analyses are above the cut in point for RNS of approximately 350 0F and 400 psig.
Adequate fire protection of equipment and circuits located inside containment required for safe shutdown has not been provided by the applicant to provide leasonable assurance that ons division will remain free of fire damage in accordance with the criteria specified above. Hose stations for manual suppression have been provided inside containment. However, due to the potontial hazard asscciated with personnel entry into containment during a plant transient, the response of the plant fire brigade may be significantly delayed, therefore no credit for manual suppression of fires inside containment during power operations is considered acceptable by the staff. In fire zone 1100 AF 113000 the applicant has .
provided a manually actuated water spray system over the non safaly related open cable trays in this zone. Both divisions of the passive residual heat exchanger (PRHR) control valves and PRHR flow transmitters are located in this zor,e in close proximity to each other. In addition. Divisions A and C of the IRWST level transmitters are located in t ,is zone with the redundant IRWST level transmitters located in the adjacent fire zone 1100 AF 11300A. There is no fire barrier or automatic suppression separating zone MOO AF 11300A from zone 1100 AF 113008. The applicant has not provided reasonable assurance that the limited manual water spray system in zone 1100 AF 113008 will maintain one division of the rarmal or passive safe shutdown capability free of fire damage.
Based on the unresolved technicalissues associated with the automatic suppression provided for equipment and circuits required for normal shutdown, capability to depressunze to the RNS cut ir, point without venting of the RCS, and the protection of safety related equipment and circuits located inside containment this item remains open.
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NRC FSER OPEN ITEM
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Response
This RAI contains several separate questions. One question is what nonsafety-related features should have automatic fire suppression, A second question is whether venting steam from the pressurizer is an acceptable method to depressurize the RCS followii g a fire. A third quevion is the adequacy of protection of safety related equipment and circuits inside containment
- 1. Which Nonsafety Related Features Should Have Automatic Fire Suppression The NRC criteria stated in this RAI is, "To enhance the survivability of the normal safe shutdown and col0 shutdown capability in the event of a 1i 9, and to reduce the reliance on the infrequently utilized sarety-related passive systems, automatic suppression shall be provided in those fire areas outside containment where a fire could damage the normal shutdown :apability, or result in a spurious operation of equpment that could result in a venting of the RCS."
The AP600 provides the following:
The NRC has stated that the safety-related systems can be the basis for fire evaluations. This is the approach that is used in the AP600 SSAR in section 9A.
SSAR secta n 7.4.1.3 describes nonsafety related features that are used to bring the AP600 to a safe shuto m and to a cold shutdown. However, this section was not written to describe the situation followit:g a fire. It does not describe the minimum set of features required. Not all support ', unctions listed in 7.4.1.3 are required including certain HVAC, chil!ed water and non-1E instruments. Theso sy:tems support front line nonsafety related systems but the front line systems can function adequately without them. A table will be added to SSAR section 9.5.1.3 to define the syt, tem capabihties used to shutdown the plant to cold conditions following a fire.
In some cases, a safsty-related capability will be used to assist the nonsafety-related caDabihties used to achieve cold shutdown. For example, the initial shutdown of the reactor will be performed by insertion of control rods which are safety related. The control rods will be used to provide this function for both the safe shutdown case evaluated in SSAR Appendix 9A and in the cold shutdown case Another example is the instrumentation used to monitor the RCS conditions to venty that it is safely shutdown. U3e of such safety-related features is acceptable.considering that their operation does not result in steaming to the containment or in loss of the RCS pressure boundary. It is also acceptable to use safety-related features to back up the CVS RCS makeup and Doration functions and its RCS pressure reduction function as long as they meet the enteria of not steaming to the containment and not losing the RCS pressure boundar',.
Acceptable means of protecting cold snutdown features from the effects of r. fire include automatic fire suppressica or separation. Separation between redundant components within a system or between different systems are both acceptable. The CVS is not separated or orovided with automatic fire suppress.on because a core makeup tank can be used to provide 280.32 s 3 iy weenmue i
i NRC FSER OPEN ITEM 11$ "y:
1 RCS makeup / boration and a ADS stage 1 valve can be used to provide a controlled, limited depressurization of the RCS. The RNS pumps and the SG PORVs are located in separate fire areas. The component cooling water system, the service water system, and the instrument air system are provided with fire suppression. Those portions of the instrumentation and control sydtem required are either separated or provided with fire suppression. The chilled water system and HVAC systems are not rcquired to support cold shutdown except for two HVAC lans that ventilate tne non-1E switchgear rooms.
In order to clarify the SSAR with respect to this question, several SSAR sections have been revised.
The revised SSAR sections include 9.5.1.3 and 9A.k.7. In addition, a table has been added to section 9.5.1.3 that lists the capabilities used to shutdown the plant to cold conditions following a fire.
- 3. Accep'3bility of Pressurizer Steam Venting for RCS Pressure Reduction The limited uso of ADS stage 1 to provide a controlled reduction cf the RCS pressure is consistent with our unoerstanding of the NRC cnteria that prohibits use of the ADS during a fire. Venting of steam from tne RCS does not cause loss of the RCS pressure boundary or an increase in the containment temperature / pressure. System level actuation of the ADS would violate this criteria.
However, the operators can open the ADS stage 1 valves individua!ly in a way that the stage 1 control valve only partly opens. Note that this valve is specifically designec to thrcitle flow. In this way, the operators can reduce tht. RCS pressure in a slow / controlled manner. When the desired reduction in RCS pressure is accomplished the partially open ADS staga one valve is ri.-closed; cepressurizing the RCS to the RNS cut-in pressure takes more than 6 minutes using the ADS stage 1 valve. The operator remains in control of the RCS pressure. A stable / visble water levelis n,aintained in the pressurizer. The RCS pressure boundary is not lost. Water :s riot discharged from the pressurizer. Steam released from the pressurizer is condensed in the IRWST and is not released to the containment. Use of the ADS stage 1 for this purpose has been added to SC,AR subsection 9.5.1.3 During several discussions with the NRC concerning spurious valve actuations. the statt stated that spunous opening of pressurizer PORV was acceptable as long as the block vaivos could be closed before control of the RCS inventory was lost. In addition, Generic Letter 31 12 says in section 4 ')
that " Power opprated relief valves may be required to reduce pressure to s':ow use c. tne high pressure injection pumps." Another example of the NRC Pccepting the use of 1:mited RCS steam venting to effect a controlled reduction in RCS pressure is found in t% Vogtle FSAR, rev. 4. TabW 9.5.1-3 of the Vogtle FSAR lists the pressurizer PORVs as a means of contrc, ling RCS pressure.
In the Open um, the NRC discusses a RELAPS analys:s Inat shows that the RCS o snd tions at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> are higher than Westinghouse has ce;ulated *430 F vs 420 F and 400 pig vs 600 psig). It is not clear that the RF. LAOS analysis uses similar inputs and usumptons 2 s the Westinghouse analysis including best 3st mate decay boch PRHR HX pert amance, and beat 'osses from the presrurizer. In any case the RCS pressure can be reduced to about 400 psig to allow RNS cut-in h,, Westag$ctse 280.32F-4
. . . . - - 1
NRC FSER OPEN ITEM maan by limited use of the ADS stage 1 valves. Note that the RNS is designed for a limited number of actuations with RCS temperatures at 420 F.
- 3. Is Protection of Safety Related Equipment and Circuits inside Containment Adequate One of the NRC concerns is the protection of the redundant PRHR HX control valves. These redundant safety related valves are fail-open air operated valves. They are located in one fire zone (11300L!) within several feet of each other. Figure 1 attached to this response shows the location of these valves. To address this concem, a noncombustible barrier has been added between the PRHR HX control valves. The barrier is made of noncombustible materials, steel or steel composites. One of the PRHR HX control valves is located close to the IRWST wall. This valve is assigned to division B. The cables for this valve are enclosed in conduit or enclosed raceways and routed up through the operating deck. Sep7 rate fire detectors are provided near each valve. The only combustibles in the area are the valves themselves and their cables. As a result, a fire that would affect these valves would start at one of the valves. The barrier protects the other valve from the initial affects of the fire. The fire detectors would alert the operators and allow them to actuate the other valve before the fire could spread and damage it. The PRHR HX control valves are qualified to operate with elevated environmental temperatures (340 F). Note that the PRHR HX flow instruments are not required to verify PRHR HX operation and have been removed from SSAR Table 9A-2. The PAMS provides this fur.ction by monitoring the RCS conditions. SSAR subsection 9A.3.1.1.8 and Table 9A 2 have been revised to reflect these changes.
For the PRHR HX to be able to cool the RCS to safe shutdown conditions in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> the IRWST gutter isolation valves must close to maintain an adequate long term PRHR HX heat sink. These redundant safety related valves are fail closed air-operated valves. They are located in one fire zone (11300A). They are separated from each other by at least 20 feet horizontal. Figure 2 attached to this response shows where these valves will be located, in addition, separate fire rieiectors are provided near eacn valve. Given the low combustible materials in this fire zone, a fire will only affect one of the valves initiall). The fire detector located at the valve that is initially affected wil' alert the operators so that they can actuate the unaffected valve before the fire can spread and damage it to the point it wiu not close. The IRWST gutter isolation valves are quahfied to ope -% with elevated env;ranmental temperaturas (340 F). SSAR subsection 9A 3.1.1.7 and Tacia 9A 2 nave been revised to reflect these changes Faure 3 shows the large vents through the rx1 of 2ones 11300A/B. Figure 4 shows cabie trays in n
(. the 113008 fire zone. Note that the safety-related cable trays are enctosed types. The nonsafety-related cab!e trays are open typos. The division B/D cab 6 tiays in zone 11399S will be enclosed J by noncombustitAs barriers made of steel or steel composite materials. The limited amount of cortioustiblas located in this zone together with its high ceiling and large volume indicate that it would not heat up ra;.;dly. Het r,ases from a fire in the ccola tray arca wiP rise away from the caole
/ trays and exit to the upper containmerit volume above operating deck. These hot gasses are impeded from crrssing over from zone 113000 to 11300A by the large I-beams that support the co., crete operating deck (see Figure 4). Some of those I beams are about 3 feet deep. If some hot t
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NRC FSER OPEN ITEM N
gas is not vented to the upper containment volume and works its way part the I-beams under the operating deck, it would be vented up to the upper containment volume through the large vents in the operating deck in zone 11300A (see Figure 3). Nnte that safety-related equipment is not located near the ceiling of these two fire zones, Also note that safety-relatad equipment located inside containment is qualified for elevated temperatures of 340 F. As a result, safety-related equipment located in 11300A or 113008 would not see elevated temperatures from a fire in the other zone that would exceed their environmental conditions.
SSAR Change:
9.5.1.3 Safety Evaluation (Fire Protection Analysis)
Subsection 9A.2.7 Safe Shutdown Evaluation Subsection 9A.3.1.1.7 Fire Zone 11300A Subsection 9A.3.1.1.8 Fire Zone 113008 Table 9A-2 Safe Shutdown Components '
ITAAC Change: /
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N9C FSER OPEN ITEM Jd.: NE Revision to SSAR 0.5.1.3 Safety Fvaluation (Fire Protoction Analysis) 9.S.1.3 Safety Evaluation (Fire Protection Analysis)
The fire protection analysts avsluates the potential for occurrence of fires within the plant and describes how hres are detected and suppressed it also confirms that the, plant can be safe'y shut
. down 'cilowing a postula'ed fire. The fire protection analysis is in Appendix 9A. I The fire protection analysis includes a set of fire area drawings and a discussion of the analysis motbodology. It also provides the following information for each file area in tha plant:
- A description of the fire area and its fire barriers, its associated fire zones, as w!! cs va
, detection and suppression capabilit:es E
e i identification of tho type quantity, and location of in sttu and anticipated transient combustible
- h. materials, and combustib% load'ng Y
- y.
- A listing of t.ufsty-related mechanical and electrical equipment E
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Fire sevchty category and equivatent duration An evaluation of fire protectsn system adequacy and tne concequences of r fire, including a
- p. dtcussicn nf the contrni and removal of smoke and hot gases, and drainage system adequacy.
I r
Far fire areas containing safety-related structures, systems, and components the fo;iowing
, information is also providad:
- k =
An eduation of fire protection Syst6m integaty. The includes a t.etermination of whether the -
E- cred'ble f ailure of a fire protection system component could cause inadvertant operation of an
$ automatic :;re suppression eystem in tre fue area, and the resulting consequences. Also
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' included is verification that n. potential single impairment of the fire protection system could incapacitate bo'h the automatic suppression system 3nd the backup manual suppress;cn system E (generally a hose station), for fire areas where both types of suppiession systems are provided.
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A safe statdown evaluation confirming tra capability to safely shut cown the reactc' and g maintain it m a safe shutdown condition following a fire w
i The salo shutdown evaluation is based upor. all components in a single fire area outside containment or any fire zona inside containment being disaDied by the fire. Success is based upon C :
the plant being able to achieve safe shutdown as discussed in SecFoo 7 4.. Safe shutdown is a 7 safe, staofe conditi6n that can be maintained indefinitef with the reactor subentical and reactor
_ coolant pressure at a small traction of its design pressure. As described in Section 7.4.1.1, i .
safety retred systems acniste this cond: tion automatically using reliable, passive pmcesses. The Ti'.
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3 n passive residual heat remuval heat exchanger transfers heat to the in-containment refueling water storage tank. Steam from this tank enters the containment which is cooled by the passive containment cooting syctem. These systems reduce the reactor temperature and pressure to less than 420* F and 600 psia in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The containment heats up to less than 225* F and less than 22 psig ouring this time. This is a safe and acceptable end state which is used to show compliance with BTP 9.51. The safe shutdown fire evaluation in Appendix 9A shows that there is sufficierit s safety-rela'.ed equipment availab!e after a fire which destroys a sir 4gle fire area outside containment or any fire 2er'o inside co'itainment, to bring the plant to this safe shutdown condition.
It should Le noted that fo: lowing most fires, that nonsafety rolated systems are expected to be available to bring the plan' to a cold shutdown for repairs. %ese syste ns ln;;ude the norme; tesdue: rcet reme4 system, coms,enent-ecebeg-weteesys",em, and the said;e weier system.
These systems-a e t'eseebsd in CP.AiW,ect; ens 5.4.7, 0.2.2, and 0.2.t . These systems are deferse in depth systems with redundant active components. Use of nonsafety-releted ,,ystems to achmete8+sku'dwa is desenbe&m-Sect;en 7.4.t.0. These systems are expected to be available because o' the use of ledundant equipment and eetwe fire protection features, including separation er automatic fitu supprAssion.
Tab e 9.51.1 lists the system capabilities that are expected to be available following a fire to bring the plant to a cold shutdown. This list does not contain the nonsafety-related support systems that are not necessary to operate to!% wing a fire. For example, chilled water cooling and non 1E instatmentation are not required inliowing a fire. Heating and ventilating are not required except *ar two fans used to ventilata the nJn-1E switchgear rooms. The fo!!owing safety related capt W 3 are used together with these nonsafety related capabilities to achieve cold shutdown:
lnsertion of contro! rods to provide reactor shutdown, lnstrumentation to monitor reactor coolant system conditions.
Operation of one core makeup tank in a natural circulation noode to provide reactor coolant makeup and boratiatr in case lho chemical and volume control system makeup is unavailable due to a fire, Manual partial openong (and closing) cf one first stage automatic depressunzation valve to provxte a control led, hmited depressuriz1 tion of the reacMir .:colant system to allow initiation of 1 tha normal residuall'est removal system in case the chemical arid volume contic! system l a. dhary spray is unavailable duo to a fire.
The u.e at these safety.related cap 1bihtues dans not result in significant plant transients. The reactor coolant system pressure boundary is maintained and containment pressure arid temperature
) conditions are not affected by tho use of these safsty-related capabilities.
11 a lens liksty, more severe fire occurs, these systems are expected to be recovered after reasonable actions are taken to uttH2e temporary connections or to perform repairs (seo subsections 9.2.2.4.5.S and 0.5.1.1.1). Recovery of thesa systams al!ows the piant to be brought to a cold shutdown for plant repairs. No credit is taken in the Appendix 9A fire evaluation 1r nonsafety related systems. As a result, fire separation is not required for these systems.
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Nsw SSAR Table 9.5.14 Table 9.5.14 Gapabilities Used To Achieve Cold Shutdown FOllowing a Fire h
h l Function System Capability Fire Protection RCS Roactivity Control
. Snart Term - Control Rods separation {
, Long Term - (1) - (1) l RCS Makeup - (1) - (1)
RCS Pressure Control
, Increase - Pressurizer heatert separation
, Decrease - Auxistary spray (2) - (2)
Decay Heat Removal - SFW pumps feeding CST water to SG fire suppression (high temperature) - SG PORV discharge to atm. separation a
Decay Heat Removal - RNS pumps circulating RCS separation (cold temperature) - CCS cooling RNS fire suppression SWS cooling CCS - fire suppression Process Monitoring RCS monitoring instruments - separation (PMS)
Non-1E Instrumentation and Control (3) separation Support Systems - Instrument Air - fire suppression Standby Diesel Generators fire suppression and ,
separation
- Non 1E AC Power (3) - separation
- Non-1E Room Ventilaticn Fans (4) separation t
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New SSAA Tabie 9.S.14 (cont.)
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\ a' Notes: (1) CVS niah.up finm the DA T,iuside? RC5 A3keup r.; batation. If I:.w C\iS is damaged by g i a tire, one CidTcan provide this croat.-':./ without heating up itm containment. >
(7) C VP aux:'iary .sy?) i revictec press:inzer pressure reduction I! lhe CVS is damage's y E. 1
- fare , o.% > ?? stago 1 valv9 L
- s.s in ? low :.acacity ttrotti'd sin!inode of .sp'::'?tioa c.tn slowy oep'essurize ths RCS witnout h.,ss of RCS p essuas t>nunac?y or hesting up of the u.. ItainYtsnt. ,
x (3) The poetic, s of the ner *ti AC power.tnr: the non- tE istrumentati~, ano cc.t os system %
i remre.1 are thcse nexted to opera:c can panoms; local controlie mitt:ciar;;(switchgear /
coobt! CWael).
(4i 'h:rtions c! the non-1E heai.ny ana vent'Uti?g systems are res+ ed to ventilate the noin c;,;<rol room, non 1E rwit~.hgea toon, and the requiredptr.ivos c? the cno 1E instrumen?? tion :v.d contrt:I system l~ee note 3).
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9A.2.7 Safe Shutdown Evaluation This sabsection desenbes the methodology for evaluation of the effects of postulated fires in each ido area on the ability of the operator to achieve a safe shutdown of the plent. The criteria and assumptions upon which the evaluation is based are described in subsection 9A.2.7.1. The safety-related features of the plant designed to provide the safe shutdown capability are described in subsection 9A.2.7.2.
As indicated in subsection 9.5.1, this evaluation is based upon satisfying the requirements of BTP CMEB 9.5-1. This basis includes using safe shutdown as defined in section 16.1 in lieu of cold shutdown wherever stated in BTP CMEB 9.5-1. The trseef-the automatic deplessurization system is not used as the method for achieving safe shutdown after a fire and spurious actuation of the o
automatic depressunzation system is avoided. The passive resictual heat removal heat exchanger is used to remove decay heat for safe ' utdown as described in suasection 7.4.1.3.
In addition the plant has enhanced capacihty to achieve cold shutdown following a fire as discussed in subsection 9.S.1. This capab:hty is not relied upon in the fire evaluation containec in Appendix EA. In adGca, o assumes an enhanced cspcb;ldy of the n ;imal residua l heat remova! system-te be-avt:1hWe4e"cw ng s-hts (see subsestcn M.7). Pass.ve res4;al heat remcval ar acimal eystaTis Grea avadable.
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NRC FSER OPEN ITEM Revision to 9A.3.1.1.7 and 9A.3.1.1,8 9A.3,1,1,7 Fire Zone 1100 AF 11300A This fire zone is comprised of the following room (s):
Room No.
11300 Maintenance floor (southem part) 11400 Maintenance floor mezzanine (southern part)
Safe Shutdown Evaluation The quantity and arrangement of the combustible materials in this fire zone, and the characteristics of the barriers that separate this zone from other fire zones are such that a fire which damages safe shutdown components in this zone does not propagate to the extent that it damages redundant safe shutdown components in another fire zone.
The quantity of combustible matenals in this fire zone is low, consisting primarily of cable insulation.
There are small concentrations of cables at the top of the zone and at several separate locations along the walls. This fire zone is physically separated from fire zones below by the maintenance floor, with a concrete thickness of more than one foot, except for openings described in the evaluation of fire zone 1100 AF 11206. This fire zone is separated from the operating deck above (fire zone 1100 AF 11500) by a ceiling with a concrete thickness of more than one foot, except for the hatches near the containment maintenance hatch, which are covered with steel grating. The walls of this fire zone are the steel containment vessel, the steel Wall of the in-containment refueling water storage tank, or walls with a concrete thickness of more than one foot, except for two designated boundaries with the adjacent portion of the maintenance floor (fire zone 1100 AF 11300B). These boundaries are approximately at the centerline o' containment, one located in the narrow annular space behind the in-containment refueling water storage tank and the other near the personnel hatch. The steam generator compartments, the refueling cavity, and the in-containment refuelir.g water storaga tank provide barriers between the two large maintenance floor fire zones.
Safe shutdown components fire zone 1$00 AF 11300A are separated from redundant safe 4
shutdown components in fire zone 1100 AF 113008 by these barriers or by a horizontal distance of more than 20 fAat with no intervening combustible or fire hazards, in addition, safety-related cables in both of these fire zones are routed in closed cable trays or conduit. minimizing the likelihood that a fire originating in a raceway of one division can propagate to a raceway of another division.
Furthermore, open-nozzle water spray suppression systems are provided for nonsafety related electrical cables routed in open cable trays in fire zone 1100 AF 113000 (there are no such cable trays in fire zons 1100 AF 11300A) providing additional assurance that a fire will not propagate between these fire zones.
Most of the smoke and hot gases from a fire in this fire zone nses through the large steel grating covered hatches between the containment maintenance hatch and the steam generator 2 T Westngh0use 2m2N2
i NRC FSER OPEN ITEM compartment into the large air space in the upper portion of containment (lire zons 1100 AF 11500).
Small quantities of smoke, especially that which has already cooled, may migrate horizontally into the adjac9nt portion c' the maintenance floor (fire zone 1100 AF 113008). The smoke and gases are cooled oy mixing with the air and by contact with structural surfaces and thus do not cause propagation of the fire beyond this fire zone. Temperature effects on the electrical cables routed high above the operating deck and passing over the large steel-grating covered hatches are not expected to be significant, but are not a concem as these are the same cables that continue into this fire tone and are assumed to be lost. Safe shutdown components listed in Table 9A 2 for the adjacent fire zones are not susceptible to damage by the diluted and cooled smoke and gases from this fire zone.
Table 9A 2 lists the safe shutdown components located in this fire zone. The passive core cooling system has two IRWST gutter isolation valves located in this zone. These valves close to divert condensate from the passive containment cooling system (on the inside of the containment shell) into the IRWST. This condensate maintains the passive residual heat removal heat exchanger heat sink for the long term. These valves are fait closed air operated valves. They are Iccated at least 20 feet apart horizontally and a fire detector is located close to each valve. Given the low combustible materials in this fire zone, a fire will only affect one of the valves initially. The fire detector located near the valve that is initially affected will alert the operators so that they can actuate the unaffected valve before ths fire can prevent operation of the second valve. These valves are qualified to operate with elevated temperatures of 340 F.
Although the consequences of a fire are expected to be very limited, a fire in this fire zone is l conservatively assumed to eventually disable all of the safe shutdown components in this fire zone.
The redundant passive core cooling system, passive containment cooling system and steam generator system safe shutdown components (listed in Table 9A 2), located in fire zones 1100 AF 11207 and 1100 AF 113008, are sufficient to perform applicable functions to achievo and maintain safe shutdown.
Tr.e primary sampling system and containment air filtration system containment isolation valves, located outside the containment fire area, are redundant to the containment isolation valves in this fire zone and are sufficient to maintain containment integrity.
The redundant _ reactor coolant system cold leg flow instrumentation located in fire zones 1100 AF 113008 and 1100 AF 11301 is sufficient to perfurm applicable functions to achieve and maintain safe shutdovm.
No fire in this zone can cause spurious actions which could cause a breach in the reactor coolant boundary or defeat safety-related decay heat removal capability or cause an increase in shutdown reactivity of the reactor.
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NRC FSER vPEN ! TEM in Wg i
9A.3.1.1.8 Fire Zone 1100 AF 11300B This fire Zone is comprised of the following room (s):
Room No.
I 11300 Maintenance floor (northern part) 11400 Maintenance floor mezzanine (northern part)
Safe Shutdown Evaluation The quantity and arrangement of the combustible materials in this fire zone, and the characteristics of the barriers that separate this zone from other fire zones are such that a fire which damages safe snutdown components in this zone does not propagate to the extent that it damages redundant safe shutdown components in another fire zons.
The quantity of combustible materials in this fire zone is low, consisting primarily of cable insulation in the termination boxes and cabic trays. There is a concentration of cables on the south side of the zone near the refuehng cavity and Small concentrations of cables at the top of the zone and at several locations along the walls, This fire zone is physically separated from fire zones below by the maintenance floor, which has a concrete thickness of more than one foot, except for access stairways and hatches. This fire zone is separated from the operating deck above (fire zone 1100 AF 11500) by a ceiling that has a concrete thickness of more than one foot, except for several openngs for an access stairway, elevator, hatches and blockouts. The walls of this tira zone are the steel containment vessel, the steel wall of the in-containment refueling water storage tank, the noncombustible enclosure for the division B and D penetrations and raceways (fire zone 1100 AF 11500), or walls with a concrete thickness of more than one foot, except for the designated boundanes with the adjacent portion of the maintenance floor, described in the evaluation of fire zone 1100 AF 11300A. There is a doorway to lower pressurizer cumpartraent (fire zone 1100 AF 11303) that is closed.
Safety related cables are routed in closed cable trays or conduit. For open cable trays, which represent the only significant in situ combustibles in this fire zone, open-nozzle water spray suppression =ystems are provided These systems are automatic Except that, to preclude inadvertent actuation, operator action is required to open the outboard containment isolai;on valv6.
These suppression systems rapidly extinguish a fire in these cable trays and prevent fire propagation to adjacent fire zones.
The use of water spray systems for the open cable trays in this fire zone limits smoke and heat gsneration. Small quantities of smoka and hot gases from a fire in this f!re zone rise through openings in the ceiling, or migrate via the large steel grating covered hatches between the containment maintenance hatch and the steam generator 2 compartment in the adjacent portions of the maintenance floor (fire zone 1100 AF 11300A), into the large air space in the upper portion of containment. They are cooled by mixing with the air and by contact with structural surfaces and thus do not cause propagation of the fire beyond this fire zone. Safe shutdown components listud 280.32F-14
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NRC FSER OPEN ITEM
- in Table 9A-2 for the adjacent fire zones are not susceptible to damage by the diluteo and cooled >
smoke and gases from this fire zone.
Tatsle 9A 2 lists 'he safe shutdown components located in this fke zons. The division A and C electrical penetrations listed in Table 9A 2 are conservatively assumed to be disabled as a result of a fire in this fire zone. The 8 and D electrical penetrations ano their cable trays routed from the electrical penetrations up to the operating deck are functionally part of fire zone 1100 AF 11500.
These two divisions art sufficient to perform applicable functions to achieve and maintain safe shutdoven.
These division 8 and D electrical penetrations and their associated raceways are protected from a fire in this fire zone by a combination of barriers, distance and fire suppression systems.
Noncombustible barriers of stool or steel-composite construction form vertical shaft (s) from the floor up to the operating deck, surrounding the division B and D penetrations and the .nssociated cable trays. The significant combustible materials in this fire zone are the nonsafety rblateo cables routed in open cable trays. These cable trays are located at least 20 feet from the division B and D penetrations and their associated raceways, and they are protected by water spray suppression systems, The passive core cooling system has-the two passive residual heat removal heat exchanger control valves which are located in this fire area. These valves are fail-open Pir ooerated valves. Tney are located within several feet of each other. The valves are separated from each other ty a noncombustible barrier of steel or steel composite materials. One of the valves is located close to the IRWST wall. This valve is assigned to division B. The cables for this valve are enclosed in conduit or enclosed raceways and routed up through the operating deck. Separate fire detectors are provided near each valve. The only combustibles in the area are the valves themselves and the? cables. A fire that would affect these V:1ives would be expected to start at ona ct the valves.
- The barrier protects the other valve from the it'stial effects of the fire. The fire detectors would a:ert the operators and allow thern to actuate the other valve before the fire could spread and damage it.
These valves are qualified to operate wth elevated temperatures of 340 F. The sisnad vehte asacciatM wth :he wane sper& tee h m;unte on the voNe opsiato: / fire h this M zone lc;;ted near the vane wAl cowe thF so!enG4 vane !G4 en 30;h the v;No fa33 spen. A tre KHhis fde ;oes thettamages the a:ecinca! pcwer supplj to ine vet.as v.nl issult in the on-cixtedee cousmg-the veNetof;;lopen Reactor coolan't system, and steam generator system instrumentation located in this fire zone are conservatively assumed to be disabled as a result of a fire in this fire zone. The redundant cassive core cooling system instrumentation, and the passive containment cooling system, reactor coolant system pressurtzer and steam generator system iristrumentation iocated in fire zones 1100 AF 11206,1100 AF 11300A,1100 AF 11301 and 1100 AF 11500 are sufficient to perform the apphcable functions to achieve and maintain safe shutdown.
280.32m s T wem muu
NRC FSER OPEN ITEM Reactor coolant system temperature instrumentation located in fire zones 1000 AF 11301 and 1000 AF 11302 are sufficient to provide the monitoring fs ,ction accomplished by thi passive residual heat removal heat exchanger flow instrumentation located in this fire zone.
[ The reactor coolant system to chemical and volume control system stop valves located in this fire L
zone are conservatively assumed to be disabled as a result of a fire in th:3 fire zone. The chemical and volume control system containment isolation valves located outside of this fire zone provide backup isolation capability to maintain the reactor coolant pressure boundary.
The redundar't reactor coolant system cold leg flow instrumentation located in fire zones 1100 AF 11300A and 1100 AF 11301 is sufficient to perform applicable functions to achieve and maintain safe shutdown.
The chemical and volume control system and the liquid radwaste system containment isolation i
valves located outside the containment fire area are redundant to the containment isolation valves inside containment in this fire zone and are sufficient to perform the applicable functions to maintain containment integrity, K_
_ The redundant steam line pre;sure instruments located in fire area 1201 AF 05 for steam L generator 1 and in fire area 1201 AF 06 for steam generator 2 are sufficient to perform the
- applicable functions to achieve and maintain safe shutdown.
The redundant core ex;t thermocouple - located in fire zone 1100 AF 11500 are suffictent to provide the applicab!e safe shutdown monitoring function. '
No fire in this zone can causs spurious actions which could cause a breach in the reactor coolant boundary or defeat safety-related decay heat removal capability or cause an increase in shutdown reactivity of the reactor.
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SAFE SHUTDOWN COMPONENTS Fire Areal tass 1E Division Fire Zone System Descrip'lon A C B D 1000 AF 01/ Core Makeup Tank B V015B V0148 1100 AF Discharge isolation Valve 11207 1000 AF01/ RNS Suction from IRWST Cont. V023 1100 AF isolation Valve 11208 Retum from CVS Cont. V061 Isolation Valve
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1000 AF 01/ PS' Containment Air Sample V008 1100 AF Cont. Isolation Valve 11300A Liquid Sample Line Cont. V010A V010B kolation Valve RCS .old Leg 2A Flow FT 103B FT-103D Cold Leg 28 Flow FT 104B FT-104D VFS Containment Purge V009 Discharge Cont. Isolation Valve VFS Containment Pur ,e inlet V004 Cont. Isolation Vt've g , 280.32F-17 1
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Table 9A 2 (Sheet 3 of 17)
SAFE SHUTDOWN COMPONENTS Fire M e d Class 1E Division Fire Zone System Description A C B D 1000 AF 01/ PXS IRWST Level LT-046 LT-048 1100 AF 11300A IRWST Gutter Isolation V130A V1308 Valve Core Makeup Tank (MT-02A)
PCS Containment Pressure PT-006 PT-008 SGS Steam Generator 2 Wide LT 014 LT-018 Range Level 1000 AF 01/ CCS Outlet Line Cont. Isolation V207 1100 AF Valve 113008 CVS Letdown Containment V045 Isolation Valve Makeup Line Cont, V091 Isolation Valve RCS Purification Stop V001 V002 Valve (RCPB)
(DS Class 1E Electrical EY-P11Z EY P27Z Penetrations Class 1E Electrical EY P12Y EY-P29Y Penetrations Class 1E Electrical EY-P13Y EY P28Y Penetrat'ons Class 1E Cable Trays Note 1 Note 1 PCS Containment Pressure PT-005 PT-007 28a32M 8 W westinghouse
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Table 9A 2 (Sheet 4 of 17)
SAFE SHuiDOWN COMPONENTS Fire Areal less ivision Fire Zone System Description A C B D l 1000 AF 01/ PXS PRHR Heat Exchanger V108B V108A 1100 AF Control Valve 113008 IRWST Level LT-045 LT-047 I Inlln I:ow IT-040A IT-0400 Core Makeup Tank 4 (MT-028)
RCS Pressurizer Pressure PT 191 A PT-191C Reference Leg TE-193A TE 193C Temperature Pressurizer Level LT 195A LT-195C PRHR Heat Exchanger TE 161 Outlet Temperature Cold Leg 1 A Flow FT 101 A FT 1010 Cold Leg 18 Flow FT 102A FT-102C Cold Leg 2A Flow FT-103A FT-103C Cold Leg 28 Flow FT 104A FT-104C SGS Steam Generator 1 LT 001 LT-003 Narrow Range Level Steam Generator 2 LT-005 LT-007 Narrow Range Level
- Steam Generator 2 Wide LT-013 LT-017 Range Level Steam Generator 1 Wide LT-011 LT-015 Range Level SG1 Steam Line Pressure PT-030 PT-032 s SG2 Steam Line Pressure PT-034 PT-036 o.
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s NAC FSER OPEN ITEM I Ouesbon 410.339F (OITS 6194) REVIS10N 1 360 pago 4 n The system description, design parumeters, and P&lDs are provided in Section 9.4.2, Table 9.4.2-1 through 9.4.2 7, and Figure 9.4.21, of the SSAR. respectively. Table 3.2 3 of the SSAR providu the dassificatica of the annex / auxiliary buildings non radioactive beating ventilation and air conditioning sys:cm (VXS) system and components, flowever, Westinghouse needs to provide the following:
- 1. updated P&lDs
- 2. 4 st.sndard safety analysis report (SSAR) figure reflecting the updated P&lDs showing major insturaentation, and sptem interactions with other systems such as supply of the chilled and hot water as provided from the i central chilled water system (VWS) and plant hot water system (VYS).
- 3. a ngure for the ancillary diesel gen rat?r room showing air supply to and from the mechanical equipr.ient room heating ventilation and air cwitioning system (HVAC) subsystem.
-4. SSAR table listing major subsystem component p:.rame'ers for the VXS including air handling units, supply ad exhaust fans, electric unit heaters and headng coib, and ancillary diesel generator room exhaust fan.
- 5. update SSAR Table 3.2 3 to include classification data for the anci!!ary diesti gtnerator som exhaust far and correct code data for filters and fes.
- 6. revise SSAR Section 9.4.2.2.2 to include descriptions for humidifiers, hot wates unit heaters, and isolation dampers.
- 7. revise SSAR Section 9.4.2.2.3.5 to state that "To replace the filters cf an air handling unit, the unit s 2 i
stopped and isolated from the duct synem by means of isolation dampers."
- 8. revise SSAR Section 9.4.2.2.1.3 to desigrate "reector switchgear" as " reactor trip switchgear."
- 9. revise SSAR Section 9.4.2.2.3.1 to state !!vt the fetwd areas in the armex building are m tintained at a positive pressure (state pressure) with respect to the adjacent areas during filter replacement mode.
- 10. revise SSAR Section 9.4.2.L2 to provide the design temperature conditions for the elevator machine room and boric acid batching room.
Response
- 1. VXS P&lDs (Rev. 5) will be forwarded to the NRC via separate correspondence.
- 2. Applicable updated SSAR figures for VXS are attached.
- 3. SSAR figure 9.4.21 (Sheet 5 of 5)is attached and has been revised to show the 14VAC servicing the ancillary diesel..
4 _ SSAR Tabit.s 9.4.2 i and 9.4.2 2 provide major subsystem component parameters for the VXS defense-in-
. depth systems. Similar information requested for nondefense in-depth components is considered to be escessive since basic inforrution is already provided in the text _ Also, there is no reason to control this nor4anse in-depth information in Tier 2. For these reasons, no change to the SSAR is necessary.
- 5. De . .aust fan for the ancillary diesel generator room provides ventilation for the diesel fuel stored in the area. The fan does not operate during operation of the ancil_lary d'esel gene.ator. " Ventilation and cooling for the room when tne ancillary diesel generators optrate is provided by means of manually operated dampers and opening doors to allow radiator discharge air to be exhausted direct to outdoors.", as noted in SSAR 9.4.2.2.1.5. For this reason, no change to the SSAR is necessary.
> 1 p 410.339F(RI) 1
NRC FSER OPEN ITEM
- 6. SSAR 9.4.2.2.2 will be updated to include descriptions for humidifiers, hot water unit heaters, and isoladon dampers.
- 7. SSAR 9.4.2.23.5 will be revised to state, "To replace the filters of an air handling unit, the unit is stopped and isolated from the duct system by means of isolation dampers.", ks requested.
- 8. SSAR 9.4.2.2.13 will be revised to designate " reactor switchgear" as " reactor trip switchgear.", as requested
- 9. The general area llVAC subsystem will maintain a slightly positive pressure during filter replacement as 3d in 'he SSAR. A positive pressure exists because the supply air flow rate is madntained greater than [
the exhaust air flow rate. He reason for maintaining a positive pressure in this subsystem is to provide additional assurance that controlled access areas such as the health physics area are maintained et a negative 6 pressure relative to other areas of the building. His is a nonsafety related function Since the health physics and hot machine shop ilVAC system normally maintains a slightly negative press,'re by design, even if the general area IIVAC subsystem is not operating, the controlled access areas would still be maintained at a negative pressure. For this reason, no change to the SSAR is necessary.
- 10. The design temperature conditions for the elevator machine room and boric acid batching room will be added to SSAR 9.4.2.1.2 as requested.
SSAR Revision:
- l. None. Ilowever, forward VXS P&lDe (Rev. 5) to the NRC via separate correspondence.
- 2. Update SSAR figure information based on attached VXS figures.
- 3. Update SS AR figure 9.4.21 (Sheet 5 of 5) to show the HVAC servicing the ancillary diesel. (Attached.)
- 4. None.
- 5. None.
- 6. Update SSAR 9.4.1.2.2 to include desc.iptions for hutaidifiers, hot water unit heaters, and isolation dampers as follows:
Just before the section entitled Shutoff, Control, Balancing and Backdraft Dampers add these three subsections:
"Humidtfier The humidyler is a packaged electric steam generator type which converts water to steam and distributes it through the supply duct system. The humidyler is performance rated in accordance wiA ARI 620 (Refsrence 13). -
4 i 410.339F(RI) 2 W Westilighouse
i NRC FSER OPEN ITEM IHot Water Unit Heaters The hot water unk heaters conskt of a fan section and hot water heating coil section factory assembled as a complete and integral unit The ank heaters are either horkontaldhcharge or sertkal downblast Ope. . The coil ratings are in acco,4ance with ANSI /ARI 410 (Reference 12).
. Isolation Dampers isolation dampers are bubble tight, single or parallel blade type. The kolation damptes have spring return actuators which fail closed on loss of riretricalpower or loss of air pressure. The kolation
, dampers are constructed, qualyled and tested in accordance with ' ANSI /AMLA 500 (Reference 14)."
- 7. Add the following to the end of the second paragraph of SSAR 9.4.2.2.3.5: "To replace theffikes of an air handling unit, the unit h stopped and koksedfrom the duct system by means of kalation dampers.*
8 Revise SSAR Section 9.4.2.2.1.3 to designate " reactor switchgear" as teactor tr!p switchgear."
- 9. None.
- 10. Revise SSAR Section 9.4.2.1.2 to add the fo:!owing design temperature conditions at the end of the Normal Operation list:
Room or Atua Temperatures PO Normal Operationi
" elevator machtne room 59 105 boric acid bauhing room 5010$*
_ - . = =
.== =
MM i
, 410.339F(RI) 3
NRC FSER OPEN ITEM i I . Note: This response deals with VXS and HVAC exhaust fans for VAS, VRS and VilS. Rese items are being i addressed in this response as the result of a meeting between Westinghouse and the NRC held on Ol/16N8 in i Rockville. Action W- Per Meeting Moutes: Action W . W will provide additional ;~igures as identified in open item. W will ensure that VXS describes the elevator machine room ventilation subsystem. W will ensure that descriptions of exhaust fans for areas where a negative pressure is required states ti,at the fans can maintain the pressure (VAS, VXS, VRS, VHS). This status was es'ablished during an NRC/W meeting in White Flint on January 15 & 16,1998, I Response: (Revision 1)
- 1. De a<lditional figures are being addressed by the response to FSER Open item 410.415F revision 1.
- 2. In the meeting, it was not clear where the SSAR described the elevator machine room ventilation subsystem.
A review of SSAI; 9.4 shows that the elevator machine room ventilation subsystem is adequately described in SSAR 9.4.2.2.3.3, Equipment Room HVAC Subsysam last paragraph under Normal Operation. It states, "A temperature controller opens the outside air intake and starts and stops the elevator machine room exhaust fan as required to maintain room design temperature conditions. A local thermostat controls the electric unit heater."
- 3. De issue was that Westinghouse did not want to provide a specific sizing value for the exhaust fans and the NRC wanted to identify that the exhaust fans where adequately sized to allow the system to maintain the
" negative pressure" when noted in the 5SAR. It wts agreed in the meeting that Westinghouse should review the SSAR and modify it where appropriate to state War the exhaust fans where adequately sized to maintain a negative pressure. Consistent with the agreement, changes to VAS, VRS, and VHS are shown below under "SSAR Revision". (Note: Since VXS does not keep the area at a negative pressure, no change to VXS is required.)
SSAR Reisiort I, See Response to FSER Open item 410.415F, revision 1.
- 2. None.
- 3. Change VAS, VRS, and VHS as shown below:
A. Re: Radiologically Controlled Area Ventilation System $'AS) 1 410,339F(RI) 4 T Westinghouse 1
NRC FSER OPEN ITEM 9.4.32.1.1 A* axillary /Annes llullding Ventilation Subsystem 2nd para. I %e two 56 percent capacity exhaust air fans shed to allow t.T.' system to maintain a negative pressure are located in the upper radiologically controlled area ventilation system equipment room at elevation 145' 9" of the avuliary buildn.g. De exhaust air ductwork is routed to minimize the spread of airborne contamination by directing the supply airflow from the low radiation access areas into the radioactive equipment and piping rooms with a greater potential for airborne radioactivity. Additionally, the exhaust air ductwork is connected to the radioactive waste drain system (WRS) sump to maintain the sump atmosphere at a negative air pressure to prevent the exfiltration of potentially contaminated air into the surrounding area. , ne exhaust air ductwork is connected to the radwaste cffluent holdup tanks to prevent the potential buildup of airborr.e radioactivity or hydropn gas within these tanks. He exhaust fans discharge the exhaust air into the plant vent for monitoring of offsite airborne radiological teleases. 9.41? l.2 Fuel lland!!ng Area Ventilation Subsystem 2nd para. I ne two 50 percent capacity exhau',t att fans sizei to allow the system to maintain a negative pressure are located in the upper radiologically controlled area ventilation systern equipment room at elevation 145' 9" of the auxiliary buildmg. De supply and exhaust ductwork is arranged to exhaust the spent fuel pool plume and to provide directiorial airflow from the rail car bay / filter storage area into the spent resin equip.nent rooms, he exhaus' fans discharge the normally unfilteied exhaust air it.to the plant vent for monitoring of offsite airborne gaseous and other radiological releases.
- 11. Re: Radwaste Buildingt HVAC System (VRS) 9.4.8.2.1 General Descriptioti 3rd para.
( l ne exhaust air system consists of two 50 percent capacity exhaust centrifugal fans si:ed to allow the system I to maintain a negative pressure, an exhaust air duct collection system, and automr. tic controls and accessories, ne airflow rates are balanced to maintain a constant exhaust design air flow through the fans. De exhaust fans are lo~ated in an equipment room on Eevation 100' 0* in the northwest corner of the radweste building. C. Re. Health Physics and Hot Machine Shop HVAC System (VHS) 9.4.11.2.1 General Dscription i The exhaust air system consists of two 100 percent capacity exhaust cen'rifugal fans shed to allow the i system to maintain a negative pressure with ductwork and automatic controls, and a separate machine shop 4 exhaust fan and high efficiency filter for exhausting from machine tools and other localized areas in the hot machine shop. The exhaust fans are located in the staging and storage area on elevation 135'-3' of the anae4 building. De machine shop exhaust fan and filter are located locally in the machine shop. De air flow rater are balanced to maintain a constant exhaust design air flow through the fans.
)
1 410.339F(RI) 5 l
NRC REQUEST FOR ADDITIONAL INFORMATION El =j r OITS: 4915 RAl: 480.873 Rev.1 Pleasc provide details of the calculations Westinghouse performed to assure that the ~ evaporation limited PCS flow is equivalent to using the actual PCS film flow with a time and elevation dependent coverage fraction. Desponse: The PCS film coverage model calculates the time and elevation dependent coverage, and the resulting evaporation rate that is input to the WGOTHIC code. The PCS film coverage model uses the actual flow rate delivered to the shell. The PCS film coverage model is described in Reference 480.873 1, Sections 7.3, 7.4, and 7.5. ClerNce8lon Question 1 RAI response does not answer question. Provide the requestod comparison.
Response
The question was based on methodology in WCAP 14407, Rev. O, Section 7. The following response is consistent with me todology in WCAP 14407, Rev.1. The flow rate applied to the shell sudace in the WGOTHIC modelis called the evaporated flow rate. Time and elevation dependent coverage is calculated with a spreadsheet, as described in Section 7.S.1.3 of WCAP 14407 Revision 1, which invo!ves an iteration with WGOTHIC on the shell heat flux. Note that for the first three hours when delivered PCS flow is at 440 gpm, coverage is assumed to be 90% of the vesselperimeter, conservatively based on the Water Distnbution Test at 220 gpm equivalent delivered flow to the extemal surface of the shell. During the first three hours, evaporation does not cause the minimum film flow rate cnterion of 120 lbtryhr ft to te reached; thus, coverage is constant until weII beyond the time that peak prussult is predicted to occur at about 20 minutes after initiation of the transient. After three hours, the evaporated flow calculation (Section 7.5.1.3 of WCAP 14407 Revision 1) ' explicitly modek time and elevation dependent coverage, with an iteration on WGOTHIC shell heat flux. The evaporated flow rate is then input to WGOTHIC along with a coverage area that assures the code will evaporate the applied ficw. Sensitivities have been performed in which WGOTHIC svaporated all the water on (a) the dome only; (b) the sidewall only; or (c) evenly split between the sidewall ar'd the dom:. The sensitivities showed that WGOTHIC calculated pressure is not sensitivs to the location that water was applied. 460.873 Rl 1 I 1
.. +
NRC REQUEST FOR ADDITIONAL INFORMATION Clarification Question 2 Received via email from D.C. Scaletti to B.Rarig 1/13/98 in the original RAI the staff requested Westinghouse provide analyses using the actual PCS flow rates not the limited evaporated flow rates to confirm that the two are identical or at least prove that the evaporeted flow model is clearly conservative. The irSues are identitled in our interim FSER provided you un 12/31/37. In the clarHication response dated 1/6/98, the requested analyses are again not provided. if Westinghouse does not provide these analyses to demonstrate that the twn results are Identical (as claimed) or that the evaporated flow mod 61 is clearly conservative, then we will apply (an as yet undetermined) penalty on the final SSAR pressure results. Both the LOCA and MSLB must be analyzed through the peak pressure period since the PCS is credited for both events.
Response
The PCS film coverage model calculate the time and elevation dependent coverage and the resulting evaporation rate that 19 input to the WOOTHIC code. The PCS film coverage model is described in Reference 480.8731, Sections 7.3,7.4, and 7.5. Sensitivity cases for the LOCA and MSLB pressure resportse were made using the actual PCS flow rate, instead of the evaporation limited PCS flow rate, in the EGOTHIC AP600 Containment Evaluation Model. Except for the PCS flow, these cases were run with the model described in Reference 440.8731. The actual PCS flow rate is greater than the evaporation limited PCS flow rate.1he ' predicted peak pressure is lower for both the LOCA and MSLB cases using actual PCS flow rate. The peak pressure for each case is listed in the table below. The transient pressure comperleon for each event is shown in Figures 480.8371 and 480.873 2. Wase Case Full PCS Ficw LOCA 43.9 psig 43.8 psig MSLB 44.8 psig 44.5 pelg amamuse 480.473 RI 2
NRC REQUEST FOR ADDmONAL INFORMATION ag increasing the PCS flow rate in the evaluation model (from the evaporation limit) reduces the predicted pressure because additional heat can be removed through sensible hen Ing of the additional flow.
References:
480.873 1 WCAP.14407, Revision 1. "WGOTHIC Application to AP600", July 1997, Westinghouse Electric. Corporation. \ a SSAR Revislan: NONE 8
~
480.873 R13
NRC REQUEST FOR ADDIT;ONAL INFORMATION
!!f "M 1
1 I _0CA Pressure Comparison i Evaporation t.imited PCS Flow Rote )
----Full PCS Flow Rote 50 ,_ i n -
# ~
.- 40 ,
en _l
? 30 a3 ,: '
a 20 _ t/) - ' ~ en . a) 10 o_ -
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~~~ ' ' ' ' ' ' ' ' ' ' '
0 0 2000 4000 60'00 8000 10000 Time (SeC) Figure 480.873-1 480,873 R1-4
., Westinghouse
NRC REQUEST FOR ADDITIONAL INFORMATION 3 VS_B ressure Comaarison Evoporotion Limited PCS Plow Rote
---- Full PCS Flow Rote 50 .
m - en : 7 'N
'- 40 _ ,, .
M _
" 30 .
o :
$ -20 ~ ,
m M . o'O U- : ' ' ' ' ' ' ' ' ' ' 0 0 . 200 400 600 800 1000 Ilme (S8C)
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Figure 480.873-2 480.873 RI 5 g , 1
FSER Open ihm [ . 650.11F, Issue C 10 Eifective Operation of Containment Sprays in 1.OCA (OITS #5973) i Revision 1 As discussed in NUREO-0933, issue C.10 addressed the effectiveness of containment sprays to remove airbome radioactive material that could be present within the containment following a LOCA. This issue was expanded to include the possible damage to equipment located within the containment due to an inadvertent actuation of the sprays. This issue was resolved by SRP Section 6.5.2,
" Containment Spray as a Fission Product Cleanup System," which references ANSI /ANS 56.3 1979, "PWR and BWR Containment Spray System Design Criteria."
In a May 28,1993, letter, Westinghouse stated that the AP600 design does not include a containment spray syster1 for removal of airborac radioactive materials in the containment. Section 15.6.5.3 of the SSAR provites the details of the accident source term and mitigation techniques for the AP600 design. Status: Since issuance of the DSER, Westinghouse has connitted to provide containment spray capability for mitiga: ion of beyond design-basis accidents. liowever, the design details have not been provided to the staff (Note: The design details have subsequently been provided by Westinghouse in draft form by letter NSD NRC 97 5329, dated September 17, 1997). Therefore, this is.ue remains open until the design is submitted to the staff and the staff has the opportunity to evaluate the design.
Response
The containment spray in the AP600 is provid-l to mitigate beyond design basis accidents, it is not credited in design basis safety analyses. Issue C.10 was revised in SSAP Revision 17. ubsection 1.9.4.2.2 to note that the AP600 does not have a safety related containna., spray system. SSAR Revision 17 includes the SSAR changes identiSed in Letter DCP/NRC1039, dated September 17,1997 to incorporate the nonsafety related containment spray. Please note that the spray header shown on a separate sheet 4 of Ogu J 9.5.1 1 in the mukup was incorporated into sheet 3 of the SSAR figure. I De water for the containment spray is from the fire protection system. The source of the water in the I fire protection system is the raw water system which provides the following water treatment before I water is delivered to the fire water storage tanks: 1) ultraviolet pretreatment to kill biological I organisms and disinfect the water,2) addition of sodiam hydroxide to maintain a pli of 5.8 to 7.5, 1 3) coagulation and clarification, and 4) filtration to remove sitt and debris. No additional additives are I added to the water for the containment spray.
- Westinghouse actions on this item are complete.
SSAR Revisions: NONE. The SSAR revisions were incorporated in SSAR Revision 17. T Westinghouse 650.11(R1) 1
a NRC FSER OPEN ITEM Question: 720.440F (OITS # 6178) Revision i Passive Containment Cooling System Flooding of the PCS annulus due to plugging of the upper annulus drains is the only PRA postulated mechanism for the failure of PCS cooling. The probability of plugging is minimized in the design by including two 100 percent drains, and a weekly surveillance of the annulus Hoor and drains to identify and eliminate debris that can potentially plug the drains. WEC has credited this surveillance in the PRA but has not incomorated it in the Technical Specifications. His is Open item 720.440F,
Response
Flooding of the PCS annulus due to plugging of the upper annulus drains is modeled in the AP600 l'RA and was assigned a failure probabihty of 1.0E-04 per demand. In the PRA model, a weekly surveillance interval was considered. The technical specifications, on the other hand, specify surveillance every two years, in addition, the specific drain configuration has changed since the Level 2 PRA was revised in September 1996. As was modeled in the PRA, Revision 8, the drains were located in the Hoor of the annulus. Ilowever, the two 100 percent drain openii,gs are now located in the side wall of the shield building with screens provided to prevent entry of small animals into the drains. The fact that the surface area of the drains is vertical makes them less susceptible to clogging due to build up of randomly collected debris, as compared to a configuration with horizontal floor drains. Thus, the drain configuration is more resistant to random plugging than a horizontal floor drain configuration, and would have a smaller failure probability. On the other hand, the technical specification surveillance interval is a factor of 100 above the assumed surveillance used in the PRA model. His change is deemed to have a negative effect compared to the PRA modeling assumption of a weekly surveillance. However,in the PRA failure probability estimate, no attempt was made to tie the estimated failure probability to the surveillance interval. A relative.y co *e value of 1.0E-04 per demand was used. There is no systematic drain plugging mechanism envisioned Thus, . de 'r.d failure rate is similar to a " shock failure" mode postulated in common cause failure (CCF) models. There ta nu need to revise this demand probabihty due to the revised drain configuration. However, the uncertainty in the failure rate increases due to the increased surveillance interval. The results of sensitivity studies performed to address this uncertainty for the PCS failure probalility are summarized below: P Mallure Probability LRF Containmer t LRF/CDF
. Effeem eness 0 0001 (base case) 1.82E-08 89.2 % 10.8 %
0 001 1.84E-08 89.1 % 10.9 % 0 01 1.97E-08 88.3 % 11.7 % 0.1 3.33E-08 80.3 % 19.7 % l.0 1.69 E-07 0.0 % 100 % 720.4@(R1)-l 1
*o oe NRC U FI OPEN ITEM From the sensitivity analysis,it can be seen that a 10 fold increase in the ITS failure probability has negligible effect on lara,; release frequency (LRF). Dus,if one looks at the upper uncertainty range of PCS failure probability with a factor of 10 increase, the LRF change would be negligible. His provides us the confidence to retam the original CCF estimate of 1.0E N per demand for the PCS drains to randomly plug.
As a conclusion, the AP600 PRA model estimate of failure for PCS horizontal drains to plug due to CCF is acceptable to represent the current AP600 design w here the vertical drains with screens to prevent small ainmals from entering the annulus drains are inspected every two years. hus, the plant risk is not affected by the design change or inspection times. Moreover, the uncertainty in the 1 RF is insensitive to a factor of 10 (even a factor of a 100) increase in the PCS failure probability, as demonstrated above. Supplemental Question Received via email from D.C. Scaletti to B.Rarig 1/8/98 in response to RAI 720.440F (OITS #6178), concerning the fatture probability of the PCS (and the TS inspection interval) based on the upper annulus drains, Westinghouse has redesigned the PCS drains. Instead of being on the upper annulus horizontal Door, they are now located above the upper annulus floor elevation. Therefore, there will be an :.ccumulated pool of water occupying the PCS air annulus region (while not stated in the submittal,I believe that the pool height will be about i foot, reducing the turning region to 5 feet in height). Westinghouse needs to address the impact of this pool (which could be as cold as 40F) on the PCS performance, the air circulation rate. In addition, the holes in the shield building need ta be addressed as they represent " leakage" into the air annulus region 'iot previously considered. Response to Supplemental Question: ne external PCS annulus drains are designed to limit the raatimum depth of the pool of water, such that the distance from the bottom of the baf0e to the top of the pool is at least five feet. De maximum pool level would be l reached at 1902 seconds based on the fact that a pool I foot deep in the annulus occupies a volume of 1902 cubic
! feet (based on a 139 foot outside diameter and a 130 foot inside Jiameter for the annulus) and the PCS initial now l rate of just under I cubic foot per second.
i l The following discussion addresses 1) the effect of the pool on unrecoverable losses in the air flow path; 2) the effect 1 of a cold pool on the temperature of air flowing through the annulus; and 3) the potential for the drain paths to affect l buoyant air Oow via leakage. i , : 1J the eJJect of the pool on unrecoverable losses in the airflow path: I
! Re cffeet of the pool on unrecoverable losses in the air Dow path is negligible since the annulus riser flow area is
! 412 square feet and the downcomer flow area is 1490 square 1,et, as compared to the entrance area between the
+
downcomer and riser of 2488 square feet with a six foot high opening and 2073 square feet with water partially blocking one foot of the annulus entry opening. De velocity through the entry area from the downcomer to the
! annulus is low and a rounded wall attachment exists at the bottom of the baffle, resulting in a form loss coefficient
! nearly equal to zero (Reference 720A40F 1. Section 4). With no pool, the downcomer-to-riser entrance area is 6 l times the iser now area and with a one foot deep pool, the entrance area is still 5 times the riser now area.
720,440F(RI) 2 L ______ _______ _ _ _ _ _
.e NRC FSER OPEN ITEM l Therefore, even with the pool present, the velocity through the entrance area is still low, and the rounded wall entry l will lead to a negligible loss through this region.
I I l 2) the effect of a cold pool on the temperature of airflowing thro::5h the annulus: l l ne presence of a cold pool and its effect on the buoyancy driven cooling of containment is negligible. Cooling by l the pool could reduce the temperature of air flowing through the annulus. De effect can be bracketed by considering cooling of the minimum air flow rate through the annulus during periods where the containment is rejecting heat from the shell outer surface. Reference 720.440F-2. Figure 4-128, shows the air mass flow rate through the annulus calculated by the Evaluation Model for a LOCA. De figure shows that the external shell has begun to significantly heat up at about the same time that the external water is credited in the model, or about 337 seconds, when the air mass flow rate is at 120 lbm/sec. De figure also shows that as the containment shell continues to increase in temperature and evaporated steam is added to the flow, the air flow increases to reach a relative.j steady value of 350 lbm/sec out to 5(WX) seconds, ne influence of a cold pool on the temperature of the air would be maximited by considering the lower value of 120 lbm/sec of air flow. l The potential cooling effect of a pool of 401- water can be conservatively estimated by assuming l a) an air velocity across the pool equal to the annulus velocity - the actual velocity across the pool is 1/5 to 1/6 of j the annulus velocity, I b) the maximum temperature difference of 120F air and 40F water. l' T = 120F l P, = 14.7 psia l l V, = 15 ft/sec I
; _ _ _ _ _ _y_ _ _m l T, = 40F Total width of pool = 4.5 feet l Using the approach recommended by Holman (Reference 720.440F-3), and assuming unit depth into the paper in
! the above figure, the,following calculation gives an upper bound cooling rate due to the pool.
i T%,,, = (12(k40)/2 = 80F Cp = 0.24 Btu /lbm degF
! p = 0.0735 lbm/cu.ft.
! p = 0.1241x10' lbm/see ft i Pr e 0.71 k = 0.015 Blu/hr ft degF 720.440F(R1)-3 i
NRC FSER OPEN ITEM v m i l Re =t (pv,L)/p = 4x10' l Since Re, for now across a flat plate is 5x10' , the heat transfer will be laminar. Since the total heat transfer coefficient for laminar now across a flat plate is equal to twice the heat transfet coef0cient at x = L, the total Nusselt number. Nuo is given by Nuo = 2Nu,,o = 2 [ 0.332 Pr" Ret " ) = 375 Pow the total average heat transfer coefficient, h, is given by Nuo = hUk = 375 Therefore, a conservatively high value for h is 1.3 Btu /hr sq.ft.-degF. Using a pool surface area of 1902 square feet and a temperature difference of 80F, the cooling rate of the pool is estimated to be less than 55 Btu /sec. To put pool cooling into perspective it is useful to compare the pool ecoling rate to the energy removal rate by the PCS, approximately 37000 Blu/sec. De pool cooling is < 0.2% of the energy de'ivered to the riser annulus by the PCS Dus, the effect of the cooling due to the annulus pool water is judged a negligible energy sink. He effect of pool cooling on the annulus air temperature and thus its buoyancy can be bracketed by considering the maximum cooling rate and the minimum air Dow: AT_ = (55 Btu /sec;/(120 lbtisec)/(0.24 Btu /lbm.degF) = 1.7F, The density effect of such a decrease in the air temperature entering the snnulus is small compared to the total density change in the riser annulus, which is due to both sensible heating of the air and the addition oflow molecular weight steam. De effect of the temperature change alone on density is related to the absolute temperature change, I which is I l- (459.7+ 118.3)/(459.7+ 120.0)]* 100 < 0.3%. Since the density change is very small and is only effective
! over a few feet at the bottom of the annulus, and since the velocity is related to the squara root of the density change l over a given height, the effect on buoyancy driven air flow is negligible.
I l It should also be noted that the cool. 40F, PCS water applied to the AP600 contamment would provide improved
} energy removal relative to the DBA Evaluation Model which assumea the PCS water is applied at 120F, I
I
? 4 the potentialfor rke drain paths to afect buoyant airflow via leakage:
- Any ambient air flow coming into the downcomer through drain paths near the bottom of the annulus would only provide additional cool air at the bottom of the annulus. Dafne leakage paths, from the downcomer to the riser at
! an elevation near the top of the downcomer, are modeled in the Evaluation Model because of the potential for short
! circuiting of the annulus flow. Because of the geometry and location of the drains, there would be no shcrt circuiting
! of the annulus now.
l It is concluded that the presence of a pool up to one foot in depth in the external PCS annulus has a negligible
! impact on PCS performance.
[ W85tingh0US8
g o-NRC . JER OPEN ITEM i
References:
I 720.440F 1 Stewart and A.T. Pieczynski," Tests of Air Flow Path for Cooling the AP600 Reactor Containment," l WCAP 13328,1992, Westinghouse Electric Company I 720.440F 2 WCAP 14407. Rev.1. "WGOTillC Application to AP600" July 1997 720.440F 3 J.P. Iloiman, llent Transfer, nird Ed., McGraw Hill Book Company. New York,1972, pp 148154. l SSAR Revision: None. PRA Revision: The following change on page 40 2 of PRA Chapter 40 Passive Containment Cooling, will be made in PRA Revision 11: (note the redline / strikeout of the changes below were provided in Rev. O response to this open item; no changes have been made as a result of this Rev, I response] l Here are two 100 percent drakns in the Doorn4 Ahe-susolos vertical wall of the shield building. Weekly l Surveillance of these drains is performed every twoyears. One drain is sufficient to prevent overflowing of the passive containment cooling system to block the air inlet. He probability, q(FC), of failure of the PC event tree node due to the drain plugging is considered to be a rare event due to the following considerations:
- Re annulus is shielded from random accumulation of debris that may potentially plug the drains.
I
- The drains are located on the shield building vertical wall above the annulus floor, and have screens to l prevent small animals from entering the drains.
t
*--Sorve !bne: i p: Armed +f:en :n;;ghron : : :: L c re:neJy :ny pbtf ia##+nt*b here are no data on this failure mode. Even if it is assumed that both drains will plug once during the plant lifetime, the failure probability would be SE-04 per week, for a 40-year plant lifetime. If S: :::kly = :ilkn :
ae :::d e- : :,i=: 6: fdb:: .e,k =i": ; perbabdity cf 0.0! per =veilkne: cpper:::i:yr4he-peobaMiity of-the fik:: ::"nuing h:: : =and "::k c;!d be 4IkO(* Based on the rarity of this failure mode and engineering judgement, a failure probability of .0001 is assigned to the PC node, given that a core damage event has occurred. q(PC) = IE 04. Figure 40-2 of PhA Chapter 40, Passive Containment Cooling, will be revised in PRA Revision 11 as shown on the next page. 720.440F(RI) 5 l l
i NHC FSER CPEN ITEM
.m- w Question: 720.441F (OITS #6179) Re 4sion ! '
Reactor Cavity Thoding System: The IRWST injection squib valves are divern from the containment recirculation squib valve.A Diversity between k these valves is specified in SSAP. Section 6.3.2.2.8.9, but the criteria for confirming that diversity has been achieved is not provided. His needs to be a&essed by ITAAC. This is Open item 720.441F.
)
I
Response
As stated ir. SSAR subsection 6.3.2.2.8.9, the IRWST injection squib valves are diserse from .ontainment recirculation squib valves because they are designed to different deugn pessures. The following dist .ssion, taken from AP600 PRA subsection 12.5.1, further explams the diversity:
"The squib valves in tne recirculation lines are normally in a different envaanment than the squib valves in the injection lines. The injection line valves have reactor coolant system pressure on one side and the prenure head of the IRWST on the other side. These valves are designed to withstand and open under this type of load.
He recirculation squib valves have the head of the IRWST on one side and the containment atmosphere pressure on the other side. These valves do not have to support the reactor coolant system pressure, nor do they have to cpen under such conditions" Thus, the IRWST injection squib valves are dezigned to withstand high pressure of approximately 2500 psig whereas the recirculation squib valves are designed for a lower pressure of approximately 150 psig. Because these two sets of squib valves are desi;,ned to withstand different design pressures, the tt.Mkness of some of the valve cornponents and tha . :e of the propellr.nt charges are different. *3ecause of these differences, the IRWST in,iection squib valves are diverse from the recirculation squib valves. Because diversity is derived from the difference in design pressures ud operating conditions, there is no need for an ITAAC. l NRC Follow-on Question (frorn Enclosure 6 of NRC letter cated January 7,1998): i l The IPWST injectionJquib valves are claimed to be diverse from the cm. aent recirculation squib valves because l they are exposed to and designed to open against different system ?ressares. A such, the thickness of scme of the
! VCve components and the size of the propellant charges are diftereot. Although the difference in the valve design
! pres ure provides some degree of diversity, other failure mechanisms could affect both v alves, such as: (1) failure
! of the valve actuatln signal or power supply, (2) maintenance or surveillance errors. particularly if maintenance is performed by the same crew, and (3) failure of the propellant charges due ta &iects in the chemical compo:iuon or environmental / aging effects, particularly if charges from the same supplier and batch ere used in both valves.
Additional mechanisms or administrative controls to minimize the po'c'itial for such ammon cause failure modes i should be identified. 720.441F(RI) 1 h.h Westnghouse 1
t 4 pc NRC FSER OPEN UEM ji! Jiii Response to follow-on question:
- 1. Common cause failure (CCF) of the actuation signal and the actuation power supply for these squib valves j is treated separately in the PRA. De possibility of such a CCF occurring is quantified in the PRA. Note that DAS actuates these valves such that CCF of the i MS does not fait either the IRWST injection er the containment recirculation squib valves. Also note that the pov- r supply for these squib valves is the Class IE de power supply.
- 2. CCF maintenance errors on squib valves are cons.dered incredible because of the simplicity of sqaib valves which require verj little / infrequent maintenance and because of the effectiveness of continuity in-sen Ne testing. Maintenance errors on AOVs or MOVs which are more complicated valves and require m re frequent and extensive maintenance are more probable than m:.intenance errors o:. squib valves.
- 3. CCF of the propellant and environmental / aging effects are also considered incredible. This concern is similar to a concern over CCF of eil MOVs cansed by bad grease or environmental / aging effects. Such basic elements of components are net considered as credible CCF and are not typically consised within traditional PRA CCF analysis.
Given the above, additional mechanisms or administrative controls to minimize the potential for such common cause failure modes is not necessary. Revisions: SSAR None. ' PRA None. ITAAC None. e L mem 720.441F(R1)-2 1 1
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