NSD-NRC-97-5001, Forwards Responses to Listed Ref 1 Comments as Partial Completion of Open Item 4970
| ML20136G166 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 03/12/1997 |
| From: | Mcintyre B WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Quay T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NSD-NRC-97-5001, NUDOCS 9703170222 | |
| Download: ML20136G166 (19) | |
Text
1 Westinghouse Energy Systems Ba 355 Electric Corporation Pittsburgh PennsyNania 15230-0355 NSD-NRC-97-5001 DCP/NRC0753 Docket No.: STN-52-003 March 12,1997 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T. R. QUAY
SUBJECT:
RESPONSE TO SOME OF THE TECHNICAL SPECIFICATION COMMENTS FROM NRC LETTER OF 12/24/96
References:
- 1. Letter from NRC to Westinghouse (Huffman to Liparulo), Initial Comments on the AP600 Revised Technical Specifications (TS), dated 12/24/96.
2.
Letter from Westinghouse to NRC, NSD-NRC-97-4984 (DCP/NRC0739),
Response to NRC Question on AP600 Technical Specifications LCO 3.0.3, dated 2/13/97.
Dear Mr. Quay:
Reference 1 provided 51 questions related to the AP600 Technical Specifications. The first comment was responded to in Reference 2 and Westinghouse is developing a revised response to the first comment based on informal feedback from the NRC.
The action to respond to Reference 1 conunents 2 through 51 was logged as open item tracking system (OITS) item 4970. Attached, as partial completion of that item, are responses to the following Reference I comments:
Comment Tnguc 2
Page B 3.0-10 typo 3
B 3.1.1 - Shutdown Margin 4
B 3.1.3 - MTC 5
LCO 3.1.4 - Rod Group Alignments 6
LCO 3.2.1 - Fo(Z) 8 LCO 3.4.1 - RCS Pressure, Temperature, and Flow DNB Limits 10 LCO 3.4.4 - RCS Loops, Modes 1 and 2 19 STS LCO 3.9.4 - Containment Penetrations (Refueling Operations)
~~
24 STS LCO 3.4.14 - RCS Pressure Isolation Valve Leakage I -
1mnn-m l
9703170222 970312 I PDR ADOCK 05200003 A
PDR,
' NSD-NRC-97-5001 i -
- DCP/NRC0753 March 12,1997 Comment Topic 25 Inservice Testing Program 30 B 3.4.11 - RCS Specific Activity
)
39 Table 3.3.1-1 (OT T/OP T and RCS T-hot and T-cold) 40.
Table 3.3.1-1 (Pressurizer Level Reference Leg Temperature Compensation) 41 Table 3.3.1-1 (Reactor Trip System Interlocks and P-17) l 45 Common Mode Failure Evaluation 49 Table 3.3.2-1 (Boron Dilution Block)
-50 New T/S for Passive Autocatalytic Recombiners These responses close, from the Westinghouse perspective, the comments listed above.
a i
Westinghouse is writing responses to the other comments provided in Reference 1. Responses to the comments received for T/S 3.7 and 3.9.2/3.9.4 will be provided by March 14 in preparaiton for a Westinghouse /NRC-SPLB telecon on March 20,1997. The remaining responses will be provided by March 21,1997.
Please review the attached responses. Please contact Robin K. Nydes (412)374-4125 with any additional comments related to the AP600 Technical Specifications.
lf ff gol
- Brian A.'McIntyre, Manager Advanced Plant Safety and Licensing jml Attachment cc: Bill Huffman, NRC (IL, IA)
Angela Chu, NRC (SL,5A)
N. J. Liparulo, Westinghouse (w/o Attachment) m
AP600 TECHNICAL SPECIFICATIONS WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 1
2)
Page B 3.0-10 There is a typo in the last paragraph in that Specification "5.5.9" should be "5.5.8."
1
Response
i Agree, this correction will be included in the next revision to the AP600 Technical Specifications.
4 3)
B 3.1.1 Shutdown Margin The LCO BASES states that for steam line break accidents,if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed " radioactive release 4
limits" (without referencing the 10CFR100 limits). In the October 11,1996, letter, Westinghouse explains that reference to 10CFR100 limits is deleted as they will not be applicable to AP600. Why are Part 100 limits not applicable to AP600, and what are the " radioactive release limits" stated in the AP600 TS BASES?
Response
The LCO section of the AP600 Technical Specification 3.1.1 Bases references SSAR Chapter 15 for the dose limits (TS Reference 2).
At the time of the last revision to the AP600 Technical Specifications (8/96), it was known that the 10 CFR 100 limits were being replaced by the NRC. Therefore, the Bases dose limits discussion of the 10 CFR 100 limits was revised to reference SSAR Chapter 15. This upproach allowed the Technical Specifications to be 4
finalized, since the new limits will be correctly specified in Chapter 15.
The Technical Specification Bases 3.1.1 LCO section will be revised, as follows, to specify 10 CFR 50.34 dose limits, consistent with the replacement of 10 CFR 100 limits announced in Federal Register 61FR65157,12/11/96.
i The SLB and the boron dilution accidents (Ref. 2) are the most limiting analyses that establish the SDM value of the LCO. For SLB accidents,if the LCO is violated, there is a potential to exceed the DNBR limit and to exceed 10 CFR 50.34, limits (Ref. 3). For the boron dilution accident, if the LCO is violated, the minimum required time assumed for operator action to terminate dilution may no longer be applicable.
4
~
AP600 TECHNICAL SPECIFICATIONS i'.
WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 2
l 4)
B 3.1.3 Moderator Temperature Coefficient (MTC)
The sentence in the BACKGROUND section (fifth paragraph) of the STS, "[If tho LCO limits are not met,] the core could violate criteria that prohibit a return to l
criticality, or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity,"is deleted in the AP600 TS. Westinghouse,in the October 11,1996, letter, states that this statement applies to the AP600 and will be restored in the l
final TS submittal. Westinghouse should ensure this commitment is met.
I l
Response
i The changes identified by Westinghouse in the October 11,1996 letter will be included in the next revision to the AP600 Technical Specifications. Consistent with the STS, the fifth paragraph will be revised to read as follows:
I l
i If the LCO limits are not met, the unit response during transients may not be' as predicted. The core could violate criteria that prohibit a return to criticality, i
or the departure from nucleate boiling ratio criteria of the approved correlation may be violated, which could lead to a loss of the fuel cladding integrity, j
i 5)
LCO 3.1.4 Rod Group Alignment Limits
- a. When more than one rod is not within the alignment limit (Condition D), the 4
Required Action specifies for operator to either (D.2.1) restore rods to within l~
the alignment limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, or (D.2.2) bring the plant to MODE 3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The BASES states that 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is a reasonable time to bring the plant from full power to MODE 3 in an orderly manner and without challenging plant system. Action D 2.1, which is an added option compared to STS, appears to be not logical. This is because if the operator tries and fails to j
restore rods to within alignment limit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, there could be i-insufficient time left to bring the plant to MODE 3 in an orderly manner.
I
- b. SR 3.1.4.3 differs from the STS in that the verification of rod drop time of each rod will be performed with T,,, less than or equal to 545'F as opposed to T,,,
- greater than or equal to 500*F in STS. The October 11,1996, Westinghouse letter explained that the basis for performing rod drop time measurement at a lower temperature is not AP600 specific, and that Wolf Creek has made a submittal for NRC approval and the WOG is pursuing this as a generic change.
The staff review of the Wolf Creek submittal has determined that no sufficient technical basis has been provided to date to warrant its acceptance. The staff is unaware of the WOG submittal for the generic change (to date). Therefore, SR 3.1.4.3 is not acceptable pending further resolution of the issue, i
4
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AP600 TECHNICAL SPECIFICATIONS WESTINCHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 3
Response
I
- a. Action D.2.1 will be deleted in the next revision to the AP600 Technical l
Specifications.
l 1
- b. Westinghouse has recommended, for AP600, specification of rod drop testing under what were found to be to the most limiting temperature conditions; j
however, since this is not acceptable, the temperature will be changed to T,,,
greater than or equal to 500 F in the next revision to the AP600 Technical
)
Specifications.
1 i
6)
LCO 3.2.1 Heat Flux Hot Channel Factor (F (Z)) (F Methodology) 9 q
The normalized F (Z) as a function of core height, i.e., K(Z). is deleted in LCO n
3.2.1. The BASES states that for the AP600, K(Z)is 1.0. In its October 11,1996, letter, Westinghouse indicates that the normalized F (Z) curve for this plant is flat j
q as it is not limited by small break LOCA results. This conclusion will be subject to confirmation upon the completion of staiT review of Chapter 15 LOCA analysis.
Response
j No response required.
8)
LCO 3.4.1 RCS Pressure, Temperature, and Flow DNB Limits
- a. LCO 3.4.1 indicates that the RCS DNB parameter LCO limits are specified in COLR. This is a deviation from Westinghouse STS. In the October 11,1996, letter, Westinghouse indicates that this change is consistent with the WOG proposal to add DNB limits to the COLR in WCAP-14483, " Generic Methodology for Expanded Core Operating Limit Report," November 1995.
This report is still under staff review. Prior to its approval, the removal of the RCS DNB parameters limits from TS to COLR is not acceptable.
- b. There is an inconsistency in the AP600 TS. Though LCO 3.4.1 indicates the RCS DNB parameters limits are specified in COLR, LCO 3.4.1 is not listed in Specification 5.6.5, Core Operating Limits Report (COLR), but is listed in Specification 5.6.6, Reactor Coolant System Pressure and Temperature Limits Report (PTLR).
l i
AP600 TECHNICAL SPECIFICATIONS WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 4
Response
i
- a. The removal of AP600 DNB parameter limits in accordance with the WOG proposal was pursued based on expected NRC acceptance, which was forecast for mid 1996. Since the WOG submittal has not been approved, LCO 3.4.1 will be revised to be consistent with the STS, specifying the RCS DNB parameter limits.
- b. Section 5.6.6, RCS PTLR, will also be revised to eliminate LCO 3.4.1 limits and the associated methodology reference.
10)
LCO 3.4.4 RCS Loops - MODES 1 and 2 When the LCO is not met, Required Action requires the operator to either (A.1) restore all four RCPs to operating conditions within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (an added option compared to STS), or (A.2) bring the plant to MODE 3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. This appears to be not logical because if the operator tries and fails to restore RCP operation within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, there would be insufficient time to bring the plant to MODE 3 in an orderly manner, i
Response
Action A.1 will be deleted, in the next revision to the AP600 Technical Specifications.
l 19)
STS LCO 3.9.4 Containment Penetrations (Refueling Operations)
The LCO for containment penetrations for refueling operation is eliminated in AP600 TS. Westinghouse should provide the bases for deleting this LCO.
Response
As stated in Westinghouse letter, NSD-NRC-96-4833, October 11,1996, Roadmap, NUREG-1431 LCO 3.9.4 has been incorporated in the AP600 Technical Specifications as LCO 3.6.8, Containment Penetrations. The LCO was moved from Chapter 9, Refueling Operations, because, for AP600, it is applicable in both MODES 5 and 6, not just MODE 6 as in NUREG 1431.
--a
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AP600 TECHNICAL SPECIFICATIONS l
WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 5
24)
STS LCO 3.4.14 RCS Pressure Isolation Valve Leakage STS LCO 3.4.14 for the RCS PIV leakage limits is eliminated in AP600 TS.Section 3.4.8 in the AP600 Technical Specifications, Revision 1 dated January 13,1994
+
contained specifications for leak testing of Reactor Coolant System pressure isolation valves (PIVs). Amendment O of these specifications dated August 1996 j
deleted these specifications. The Chemical Volume and Control System (CVS),
Normal Residual Heat Removal System (RNS), Primary Sampling System (PSS),
3 and the Liquid Radwaste System (WLS) all contain PIVs. The portions of the CVS, PSS, and WLS outside of their respective outside isolation valves are all non-safety systems and therefore non-seismic systems. In addition, the low pressure portions of these three systems has not been designed to the staff's position discussed in the DSER, Section 3.9.3.1, "AP600 Design Criteria for ISLOCA" (Reference DSER Open Item 20.314 [OITS 1514] for a discussion of this issue). If the PIVs in the CVS, PSS, and WLS are not leak tested, gross failure of the low pressure, non-seismic portions of these systems should be assumed and radiological consequences
)
of PIV excessive leakage should be evaluated. Even if the resolution of DSER Open' Item 20.3-14 results in the implementation of the criteria in DSER Section 3.9.3.1, i.e., the low pressure portion is designed to an equivalent RCS pressure, a failure resulting from a seismic event still has to be assumed because these systems are non-seismic. If the evaluation discussed above is not performed, reinstate the specifications for leak testing of RCS PIVs into the Technical Specifications, unless the low pressure portion of these systems are designed to the equivalent RCS pressure, and are reclassified as Seismic Category I.
Response
The scope of the NUREG-1431 (STS) B 3.4.14, RCS Pressure Isolation Valve (PIV)
Leakage, LCO is described in the Bases Background section:
"The basis for this LCO is the 1975 NRC " Reactor Safety Study" (Ref. 4) that identified potentialintersystem LOCAs as a significant contributor to the risk of core melt. A subsequent study (Ref. 5) evaluated various PIV configurations to determine the probability of intersystem LOCAs."
NUREG-1431:
Ref. 4 -
WASH-1400 (NUREG 75/014), Appendix V, October 1975.
Ref. 5 -
NUREG-0677, May 1980.
A systematic evaluation of the AP600 systems that interface with the RCS has been performed to demonstrate that the design of the systems meets the ISLOCA acceptance criteria (WCAP-14425, Evaluation of the AP600 Conformance to Inter-System Loss-of Coolant Accident Acceptance Criteria, July 1995). The report conclusion (Section 4.0) states the following:
"The AP600 has incorporated various design features to address ISLOCA challenges. These design features have resulted in the very low AP600 core damage frequency for ISLOCA compared with that of current plants. These design features are primarily associated with the RNS and are discussed in detail in Section 3 of this report as well as SSAR Section 5.4.7. This report was
AP600 TECHNICAL SPECIFICATIONS WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 6
j prepared to document the comprehensive systematic evaluation of the AP600 design for conformance to the ISLOCA acceptance criteria in the various referenced NRC documents. As a result of this study, additional design l!
features have been incorporated in the AP600 design and are documented in the AP600 SSAR...."
1 An RCS PIV leakage specification is not required for AP600, since the risk basis for having the LCO stated in NUREG-1431 is not met by AP600. The AP600 PRA (Chapter 2, Internal Initiating Events) demonstrated that none of the potential intersystem LOCAs are significant contributors to risk of core melt.
1 25)
Inservice Testing Program All the passive core cooling system LCOs, the passive containment cooling system LCO, and the main control room habitability system LCO specify system i
performance surveillances in accordance with the " Inservice Testing Frogram."
The AP600 Inservice Testing Program is discussed in AP600 TS Section 5.5.4 and will be consistent the with Section XI of the ASME Boiler and Pressure Vessel Code.. The ASME Section XI IST program does not address any system level performance requirements related to the AP600 and is not applicable to these surveillances. Detailed system performance surveillance testing criteria should be included in the AP600 TS.
l The staff notes that Table 3.9-17 of the AP600 SSAR provides " System Level Inservice Testing Requirements." Westinghouse should explain the relationship of
-this table to the AP600 TS and how these system level inservice testing requirements were established (i.e., provide the bases) and how the requirements 1
i[
in the table are intended to be implemented. The staff also notes that some of the requirements are not consistent with previous NRC positions. For example, the -
t main control room pressurization testing specified in Note 8 of Table 3.917 is not in accordance to the policy position defined in SECY-95-132, 1
1
Response
The NUREG-1431 precedents for this approach include removal of complex
'surveillances to programs such as the Ventilation Filter Testing Program and the Steam Generator Tube Surveillance Program. In these cases the requirement to perform the surveillance remains in the Technical Specifications but the procedural details are outside of the Technical Specifications.
For AP600 the requirement to perform the system level testing is specified in the Technical Specifications and the detailed requirements are specified in the SSAR Section 3.9, System Level Inservice Testing Requirements.
The purpose of the surveillances which specify System Level Inservice Testing Requirements is to periodically verify that the system will perform its safety-related function. Therefore, the requirements are based on the safety function.
In order to clarify that the SSAR System Level Inservice Testing Requirements w
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AP600 TECHNICAL SPECIFICATIONS WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 7
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apply, Technical Specification Section 5.5.4 will be revised to reference SSAR Section 3.9.6 and Table 3.917 as follows:
Selected system level inservice testing requirements specified in SSAR e.
Section 3.9.6 and Table 3.917 shall apply when specified by individual Surveillance Requirements.
SECY-95-132 The SECY-95-132 position on periodic main control room pressurization testing is consistent with Table 3.9 both specify pressurization testing each refueling cycle.
The initial (ITAAC) control room pressurization preservice testing discussed in SECY-95-132 is separate from the periodic system level inservice testing 1
requirements listed in SSAR Table 3.9-17. The one time, preservice 72-hour capacity testing requirements are specified separately in the Certified Design Manual, AP600 Document Number GW-GL-030, Rev. 2, dated 10/31/96, Table 2.2.5-2.
No additional Technical Specification or SSAR changes are needed to specify control room habitability preservice testing.
30)
B.3.4.11, RCS Specific Activity In TS bases B.3.4.11, RCS Specific Activity, Westinghouse states that "the LCO limits are established to be consistent with the design basis fuel defect level of 0.25 percent..." as the basis for the LCO limits. This statement implies that the staff accepts 0.25 percent fuel defect level, which is contrary to the staff position in Chapter 11 of the DSER. Westinghouse should reword this sentence to avoid misunderstanding.
The staff accepts the assumption of one percent fuel defect, not 0.25 percent, for the safety analysis in SSAR Chapter 11. For iodines and noble gases, the staff accepts the limit of 0.4 pCi/gm as an exception based on this limit being subject to TS control. The staff does not accept 0.25 percent fuel defect level. Therefore, Westinghouse can not use it as the basis for establishing LCO limits.
Response
The Specification 3.4.11 Bases will be revised to delete " design basis" from the second sentence of the second paragraph of the Background, consistent with the resolution of OITS 1171 (DSER 11.2-5). The sentence will be revised to read as follows:
The LCO limits are established to be consistent with a fuel defect level of 0.25 percent and to ensure that plant operation remains within the conditions assumed for shielding and DBA release analyses.
I.
AP600 TECHNICAL SPECIFICATIONS WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 8
1 39)
Table 3.3.11, Reactor Trip System Instrumentation
]
Two of the reactor trip functions - Overtemperature Delta T and Overpower Delta T, require inputs from reactor coolant inlet and outlet temperatures. However, RCS T-cold and T hot were not listed in Table 3.3.1-1, " Reactor Trip System Instrumentation." Table 3.3.1-1 should be updated to include these two inputs.
j
Response
]
The AP600 Technical Specifications include OTAT and OPAT Functions (Functions i
6 and 7) identical to the STS. Neither AP600 or the STS include the RCS T-cold or T-hot inputs in Table 3.3.1-1 suggested in the NRC question above.
1 Specification of the temperature input channels is not needed since they must be operable as part of the operability of the OTAT and OPAT Functions.
No change is needed.
40)
Table 3.3.1 1, Reactor Trip System Instrumentation i
The pressurizer level reference leg temperature compensation input provides density compensation in the pressurizer high water level protection function.
i Table 3.3.11 should be updated to include this input.
i
Response
The AP600 Technical Specifications include a Pressurizer Level Function (Function
- 9) identical to the STS. Neither AP600 or the STS include the reference leg temperature compensation input in Table 3.3.11 suggested in the NRC question above.
Specification of the temperature compensation input is not needed since it must be operable as part of the operability of the Pressurizer Level Function.
No change is needed.
41)
Table 3.3.1 1, Reactor Trip System Instrumentation Table 3.3.1-1 item 16, " Reactor Trip System Interlocks" should include P 17,
" Power Range Nuclear Power Negative Rate Below Setpoint Blocks Automatic Rod Withdrawal," to be consistent with design requirements.
Response
P-17 was included in draft versions of the AP600 Technical Specifications.
However, it was determined, while trying to specify the safety analysis basis for
0 AP600 TECHNICAL SPECIFICATIONS WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 9
1 the P 17 setpoint, that no credit was taken for the automatic rod withdrawal block for dropped rod protection and that there was no basis for a setpoint.
Although the I&C design provides for the P 17 block, there is no safety analysis reason for including the function in the Technical Specifications.
45)
Section 5.0, Administrative Controls Because of the increased potential of common mode failure in software based systems, NRC has in past ALWR reviews taken the position that administrative controls are needed to raise the level of awareness that I&C failures must be carefully evaluated to determine if the root causes are hardware or software j
failures. Failures traced to software must be carefully considered for their common mode failure potential.
l l
The following proposed wording may be used for " Common Mode Failure Evaluation Program" under Administrative Control in Section 5.0 to address the above concern.
l "This program provides controls to ensure that appropriate software and hardware i
evaluation procedures are established to protect the plant from common mode failure, and to ensure that redundant system capability is not adversely affected.
i This program shall evaluate the cause ofinoperability, the affected components, and the plans and schedule for completing proposed remedial actions. If a determination is made that a common mode failure exists within independent i
channels or independent systems credited to provide functions controlled by l
Technical Specifications, then a Special Report shall be submitted to NRC within 30 days. The report shall include a description of the cause of the failure, the i
affected components, and plans and schedule for completing proposed remedial 1
activities."
l
Response
j l
The Westinghouse position on the common mode failure program has been j
previously provided in response to RAI 420.101 and in at least one meeting (August 10 and 11,1993).
In summary, the actions which would be required by the proposed program are already included in existing regulatory requirements covering operability determinations, QA program, and reporting requirements specified in 10CFR50.36(c)(2),10CFR50 Appendix B, Criteria XVI, Corrective Action, i
10CFR50.72,10CFR50.73, and 10CFR21. Westinghouse considers the existmg i
regulatory requirements to be more than adequate assurance that common mode failures will be thoroughly investigated to establish the cause, that remedial action will be implemented and that the failure and corrective actions will be reported to the NRC in a timely manner.
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A,P600 TECHNICAL SPECIFICATIONS WESTINGHOUSE RESPONSES TO NRC QUESTIONS AND COMMENTS 10 i
49)
Table 3.3.21, Engineered Safeguard Actuation System Instrumentation Table 3.3.21 on ESFAS Instrumentation shows the Boron Dilution Block (Item 15) nominal trip setpoint at 1.6 X the source range flux in a 50 minute period. SSAR Section 15.4.6.2.2 implies that the setpoint is based on a 10 minute period rather than 50 minute. This discrepancy should be resolved, i
Response
The AP600 Technical Specification setpoint (1.6 X in a 50 minute period) bounds the SSAR setpoint (1.6 X in a 10 minute period). The specified time period (50 i
minutes) is the period of flux level computer data memory. The current flux level is compared to each prior reading in the computer memory to determine if the flux level has increased by 1.6 or greater.
The increase in the time window from 10 minutes to 50 minutes increases the sensitivity of the system to respond to lower dilution flow rates but may also increase the occurrence of spurious actuation signals.
No Technical Specification change is required.
50)
Passive Autocatalytic Recombiners (PARS)
Westinghouse should justify why passive autocatalytic recombiners (PARS) are not subject to any TS LCO. The PARS should have operability and surveillance requirements to ensure their availability for hydrogen control.
Response
A new AP600 Technical Specification LCO will be added to specify requirements for the PARS. The surveillances will specify visual inspection and performance testing in accordance with the Inservice Testing Program.
The Section 5.5.4 Technical Specification Inservice Testing Program, will require PAR testing each 24 months in accordance with the SSAR Table 3.9-17, System Level Inservice Testing Requirements.
See also W response to NRC question #25.
A draft of new LCO and Bases 3.4.10, Passive Autocatalytic Hydrogen Recombiners, is attached.
Passive Autocatalytic Hydrogen Recombiners 3.6.10 3.6 CONTAINMENT SYSTEMS 3.6.10 Passive Autocatalytic Hydrogen Recombiners LCO 3.6.10 Two passive autocatalytic recombiners (PARS) shall be OPERABLE.
APPLICABILITY:
MODES 1 and 2.
ACTIONS CONDITION REQUIRED ACTION COMPLET10N TIME i
A.
One PAR inoperable.
A.1
NOTE------
LCO 3.0.4 is not applicable.
Restore one PAR to OPERABLE 7 days status.
B.
Two PARS inoperable.
B.1 Verify by administrative means 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> that the hydrogen control function is maintained.
AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> j
thereafter AND B.2 Restore one PAR to OPERABLE 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> status.
C.
Required Action and 0.1 Be in MODE 3.
8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> associated Completion Time not met.
bAP600 3.61 03/97 Draft 14030410 per431297
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-,._,,_-,,m,_
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Passive Autocatalytic Hydrogen Recombiners 3.6.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 1
SR 3.6.10.1 Visually examine each PAR enclosure and ensure In accordance with there is no obstruction or blockage of the inlets or the Inserv'ce outlets in accordance with the Inservice Testing Testing Program Program.
)
i SR 3.6.10.2 Perform a surveillance bench test on a specimen in accordance with removed from each PAR in accordance with the the Inservice Inservice Testing Program.
Testing Program bAP600 3.62 03/97 Draft 16030610 pat 41297
9 Passive Autocatalytic Hydrogen Recombiners B 3.6.10 B 3.6 CONTAINMENT SYSTEMS B 3.6.10 Passive Autocatalytic Hydrogen Recombiners BASES l
BACKGROUND The function of the hydrogen passive autocatalytic recombiners (PARS)is to l
eliminate the potential breach of containment due to an uncontrolled hydrogen-1 oxygen reaction.
]
PARS are required to reduce the hydrogen concentration in the containment following a loss-of-coolant accident (LOCA). The PARS accomplish this by recombining hydrogen and oxygen to form water vapor. The PARS are self l
initiated in the presence of hydrogen.
Two 100 percent capacity independent passive autocatalytic recombiners are provided. The PARS are passive devices which contain no moving parts and do not need electrical power or any other support system. Recombination is j
accomplished by the attraction of oxygen and hydrogen molecules to the surface of the catalyst. The two gases are combined to form water vapor via an 4
exothermic reaction. The heat produced by the reaction causes the air to rise 1
within the enclosure by natural convection. As it rises, replacement air is drawn into the PAR through the bottom, and is exhausted through the chimney where the i
hot gases mix with the containment atmosphere. The device is a molecular j
diffusion filter, not a fixed bod particle filter, and thus the open flow channels are not susceptible to fouling. A single PAR is capable of maintaining the hydrogen l
concentration in containment below the 4.0 volume percent (v/o) flammability limit following a DBA. Since PARS are not subject to single failure, the second PAR is installed as a spare and provides the ability to continue operations for a limited l
time, in the event one PAR is declared inoperable. The PARS will operate following any accident which results in hydrogen generation, independent of the 1
p availability of offsite or onsite power.-
i l
APPLICABLE The PARS provide for the capability of controlling the bulk hydrogen 1-SAFETY ANALYSES in containment to less than the lower flammable concentration of 4.0 v/o following i
a DBA. This control would prevent an uncontrolled hydrogen bum, thus ensuring j
the containment pressure and temperature assumed in the analysis are not exceeded. The limiting DBA relative to hydrogen generation is a LOCA.
i (continued)
- HAP 600 83.61 03/97 Draft t4030610 per431297 1
1
,,..,r
.----.---r-
Passive Autocatalytic Hydrogen Recombiners B 3.6.10 B SES (continued)
APPLICABLE Hydrogen may accumulate in containment following a LOCA as a result of:
SAFETY ANALYSES (continued) a.
A metal-steam reaction between the zirconium fuel rod cladding and the reactor coolant; b.
Radiolytic decomposition of water in the Reactor Coolant System (RCS) and the containment sump; c.
Hydrogen in the RCS at the time of the LOCA (i.e., hydrogen dissolved in the reactor coolant and hydrogen gas in the pressurizer vapor space); or d.
Corrosion of meals exposed to the post accident environment.
To evaluate the poteniial for hydrogen accumulation in containment following a LOCA, the hydrogen generation is calculated as a function of time following the initiation of the accident. Conservative assumptions are used to maximize the amount of hydrogen calculated. As such, the PARS are designed to control an amount of hydrogen generation in containment considerably in excess of the amount that would be expected from the limiting DBA LOCA.
1 Based on the conservative assumptions used to calculate the hydrogen concentration versus time after a LOCA, the hydrogen concentration in the primary containment would reach 3.5 v/o about 20 days after the LOCA and 4.0 v/o about 8 days later if no recombiner was functioning (Ref. 2).
The PARS are designed such that, with the conservatively calculated hydrogen generation rates discussed above, a single PAR is capable of limiting the peak hydrogen concentration in containment to less than 4.0 v/o (Ref.1)
The PARS satisfy Criterion 3 of the NRC Policy Statement.
a LCO Two PARS must be OPERABLE. Since the PARS are not subject to a worst case single active failure, one PAR is, in effect, an installed spare, thus allowing continued operation in accordance with Action A, in the event one PAR is determined to be inoperable.
(continued) h AP600 83.62 03/97 Drc'1ft 16030610 per431297 a
._..__._._.___._.___._.m
-._.m-I Passive Autocatalytic Hydrogen Recombiners B 3.6.10 I
BASES (continutd)
J LCO Operation with at least one PAR ensures that the post-LOCA
{
(continued) hydrogen concentration can be prevented from exceeding the flammability limit.
1 4
1 APPLICABiUTY In MODES 1 and 2, the PARS are required te control the hydrogen concentration within containment below its flammability limit of 4.0 v/o.
[
In MODES 3 and 4, both the hydrogen production rate and the total hydrogen produced after a LOCA would be less than that calculated for the DBA LOCA.
Also, because of the limited time in these MODES, the probability of an accident requiring the PARS is low. Therefore, the PARS are not required in MODE 3 or 4.
i in MODES 5 and 6, the probability and consequences of a LOCA are low due to the pressure and temperature limitations in these MODES. Therefore, PARS are not required in these MODES.
. ACTIONS
&1 With one PAR inoperable, the inoperable PA.9 must be rastored to OPERABLE status within 7 days. In this condition, the rt,maining OPERABLE. PAR is capable of performing 100% of the hydrogen control function. The 7 day Completion Time is based on the ability to perform the safety function and that no single failure of the remaining PAR is postulated. The Completion Time is further supported by the low probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit, and the length of time after the event that operator action would be required to prevent hydrogen accumulation from exceeding this limit.
Required Action A.1 is modified by a Note which states the provisions of LCO 3.0.4 are not applicable. As a result, a MODE change is allowed when one PAR is inoperable. This allowance is provided because the remaining OPERABLE PAR is fully capable of performing the safety function, the probability of the occurrence of a LOCA that would generate hydrogen in amounts capable of exceeding the flammability limit is low, the remaining OPERABLE PAR is not (continued)
HAP 600 B 3.6-3 03/97 Draft 18C10610 per431297 F
.~
q Passive Autocatalytic Hydrogen Recombiners f
B 3.6.10
]
BASES ACTIONS A:1 (continued) subject to a single failure and the length of time after a postulated LOCA be'.ne
)
operator action would be required to prevent exceeding the flammability limit.
t-f-
B.1 and B.2 V.';th two PARS inoperable, the ability to perform the hydrogen control function via altemate capabilities must be verified by administrative means within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The I
altemate hydrogen control capabilities are provided by the containment Hydrogen Ignitor System. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Completion Time allows a reasonablo period of time 4
to verify that a loss of hydrogen control function does not exist. In addition, the l
attemate hydrogen control system capability must be verified once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1
thereafter to ensure its continued availability. Both the initial verification and all subsequent verifications may be performed as an administrative check, by examining logs or other information to determine the availability of the attemate hydrogen control system. It does not mean to perform the Surveillances needed t
1 L
to demonstrate OPERABILITY of the altemate hydrogen control system. If the l
ability to prom the hydrogen control function is maintained, continued operation L
is. permitted i.a two PARS inoperable for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is a reasonable time to allow two PARS to be inoperable because the hydrogen control function is maintained and because of the low probability of the occurrence of a LOCA that would generate hydrogen in the amounts capable of exceeding the flammability limit.
i h
.C._:1 If the inoperable PAR (s) cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The allowed Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems.
(continued)
HAP 600 B3.6-4 03/97 Draft 16030A'0 par 431297 e
Passive Autocatalytic Hydrogen Recombin:rs B 3.6.10 BASES (continued)
SURVEILLANCE SR 3.6.10.1 REQUIREMENTS This Surveillance Requirement ensures there are no physical obstructions to air flow that could affect recombiner operation. A visualinspection of each PAR is i
sufficient to verify that there is no obstruction or blockage of the inlets or outlets.
The Surveillance Frequency is in accordance with the Inservice Testing Program.
SR 3.6.10.2 This Surveillance Requirement requires removal and testing of a specimen sample from each of the PARS, and it subjects the sample to bench tests to confirm continued satisfactory performance. The bench test measures the air temperature 4
increase due to exposure of the catalyst to a known air / hydrogen sample, and it demonstrates that the catalyst continues to be operable. The Surveillance Frequency is in accordance with the Inservice Testing Program and takes in to account the test data available for PARS, and the low probability of catalyst poisoning.
REFERENCES 1.
AP600 SSAR, Section 6.2.4, ' Containment Hydrogen Control Systems.'
2.
Regulatory Guide 1.7, Rev.1.
bAP600 B 3.6-5 03/97 Draft 160306f 0 par 431297