NSD-NRC-97-5278, Forwards Responses to NRC Reactor Sys Branch RAIs on AP600 Ts,Including Resultant Ssar/Ts Markups

From kanterella
(Redirected from ML20198G584)
Jump to navigation Jump to search
Forwards Responses to NRC Reactor Sys Branch RAIs on AP600 Ts,Including Resultant Ssar/Ts Markups
ML20198G584
Person / Time
Site: 05200003
Issue date: 08/27/1997
From: Mcintyre B
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To: Quay T
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20198G588 List:
References
NSD-NRC-97-5278, NUDOCS 9709040163
Download: ML20198G584 (12)
v • d • e


Text

4 O

Energy Systems k=355 Westintiouse Pmstugh Pennsylvama 15230 0355 Electric Corporation DCP/NRC0993 NSD-NRC 97 5278 Docket No.: 52-003

- August 27,1997 Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: T. R. QUAY

SUBJECT:

RESPONSES TO REACTOR SYSTEMS BRANCil RAls 440.667 TilROUGli 440.674 ON Tile AP600 TECilNICAL SPECIFICATIONS

References:

1. Letter from NRC to Westinghouse, " Comments and Requests for Additional I.7 formation (RAls) on Advanced Markup of the AP600 Technical Specifications (TS)," dated July 10, 1997.
2. Letter from Westinghouse to NRC, IX'P/NRC0909, " Advance Markup of the AP600 Technical Specifications," dated June 18,1997.

Dear Mr. Quay:

Reference i provided NRC reactor systems branch Requests for Additional Information (RAls) 440.667 through 440.674 on the Technical Specifications (TS) submitted in Reference 2. Attached are three copies of the responses to those RAls for NRC review. These RAI responses include resultant SSAR/TS markups, which are made on the Reference 2 TS markups.

With this transmittal, the Westinghouse s%tus for Open item Tracking System item 5579 is changed to

" Action N" for NRC review of the attached RAI responses.

Please contact P.obin K. Nydes at 412 374-4125 if you have any questions related to this transmittal or the AP600 Technical Specifications.

  • l Brian A. .McIntyre, Manager Advanced Plant Safety and Licensmg M [

jml

_ Attachment _

cc: W. C. IlulTman, NRC (w/ Attachment)

A. T. Chu, NRC (w/ Attachment)

N. J. Liparulo, Westinghouse (w/o Attachment) lll $!,$,h, 9709040163 970827 PDR ADOCK 052OOOOJ E PDR L _______________________________J

)

RESPONSES TO NRC REQUEST FOR ADDITIONAL. INFORMATION g.- t i

i Question 440.674 1RWST surveillance 3.5.6.3 requires verification of the IRWST boron concentration within the acceptable limit once per 31 days, which is inconsistent with 7 days specified in STS. He justification provided in

" Explanation of CT/SF Value" is not acceptable, his SR should be revised with a frequency of 7 days.

Response

As presented in letter DCP/NRC0891 of 6/6/97, there are design differences between the AP6001RWST and the RWST design used for the STS. These design differences include location of the tank inside containment and the use of a larger tank (20 to 60% larger). Location of the tank inside containment makes it less likely that water will be added to the tank, diluting the boron solution.

Two ways of changing the IRWST boron concentration have been identified, one is the addition of dilute water to the IRWST and the other is recirculation of the IRWST with a dilute water volume. The only IRWST recirculation paths are with the RNS and the SFS. The RNS can not change the IRWST buron corcentration by recirculation because it is not connected to another tank and the volume of its pipes, pumps, and HX is insignificant with respect to the IRWST. He spent fuel pool is a large volume however its boron concentration is the same as the IRWST, so recirculation with it can not reduce the IRWST below its limits.

It is possible .a dilute the IRWST boron concentration by adding dilute water to the tank. Adding enough dilute water to increase the IRWST volume by 3 % (~ 15,000 gal) would reduce the IRWST boron concentration by about 75 ppm. De IRWST boron concentration limits are 2600 ppm to 2900 ppm.

Such a small change in boron concentration (- 3 %) is insignificant with respect to the shutdown margins provided. SR 3.5.6.3 requires the IRWST boron concentration to be verified whenever its water volume has been increased by 15,000 gal. The STS( SR 3.5.4.3) do not have such a requirement.

Since the only credible way of significantly reducing the IRWST boron concentration is to add dilute water to the IRWST, monitoring the IRWST water volume provides an effective means of preventing a significant boron reduction. It is proposed that the IRWST water volume surveillance be made every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in order to reduce the possibility of the boroa concentration being reduced below its limits; the STS (SR 3.5.4.2) only require the IRWST level be surveyed every 7 days. Increasing the IRWST volume by 15.000 gal in 24 hod'rs would require an average flow of 10 gpm which will be readily detectable by other means such as RCS leakage or CVS makeup measurements.

Providing surveillance of the IRWST volume every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> together with the requirement to verify the boron concentration after a volume increase of 3% provides high confidence that the IRWST boron concentration will not be reduced to below its limits. Rese additional surveillances justify the use of a 31 day surveillance of the IRWST boron concentration.

SSAR Revision: None.

440.667 through 440.674, page 8 T Westinghouse

' L RESPONSE:S TO NRC REQUEST FOR ADDITIONAL INFORMATION

  1. ~ ni 7  ::1

- Use of stainless steel construction eliminates corrosion products.

- Good water quality is maintained such that there are fe.v particles.

- Tanks are normally closed which prevents the addition of debris.

- Technical Speci6 cation surveillance SR 3.5.6.10 requires inspection of the IRWST and containment recirculation screens every refueling. 1

- Combined License applicants referercing the AP600 are required to have a containment cleanliness program (see SSAR subsection 6.3.8.1) to prevent debris from being left in the containment following refueling or maintenance outages.

- Opening of the system for inspection or repairs is controlled by work procedures which maintain the cleanliness of the system, including the control of parts / tools to prevent them from being left in the system.

c. The name of this table is different from the table that identifies the ASME component level inservice testing (" Valve Inservice Test Requirements" vs " System Level Insetvice Test Requirements"). To provide additional differentiation, the name of the system level table will be changed to " System Level Operability Test Requirements" The Technical Specification references to this table will be revised to reflect this change, see attached markups.

SSAR Revision: See attached markups.

Referewe: 440.667-1: NUREG-1431.

Question 440,668 Surveillance Requirements should be added to verify the automatic response of all passive core cooling system automatic actuation valves that must change position in response to an engineered safety feature actuation signals to perform a safety function (similar to ECCS surveillance requirement in Westinghouse standard technical specifications, SR 3.5.2.5). The SR should demonstrate that each automatic valve responds and moves to the correct position on an actual or simulated actuation signal. This includes SR 3.5.2 for the CMT outlet isolation valves, SR 3.5.4 for PRHR air-operated outlet isolation valves, SR 3.5.6 for IRWST motor-operated recirculation isolation valves.

Response

The ESF Actuation Logic Test su.veillance (SR 3.3.2.2) together with the Actuation Device Test surveillance (SR 3.3.2.7 / 8 / 9) provides overlap testing that replaces the need for a separate surveillance as required i i standard Technical Specifications. This overlap is a result of the AP600 I&C design wFich is different than the I&C designs for which the STS were developed. During the ESF Actuation Logic Test, system level actuation signals are generated in the ESF cabinets and are sent to the component level logic cabinets. The automatic tester that performs the test only allows the system level signal to actuate 1 of the 3 redundant logic cabinet logic sets. Since the logic sets actuate components on a 2 out of 3 440.667 through 440.674, page 2 W85tiligl10158 1

D-RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION -

?

' Response to Technical Specification RAls 440.667 through 440.674 (Open Item Tracking System item 5579)

Question 440.667 Technical specification surveillance requirements SR 3.5.1.6,' 3.5.2.7, 3.5.4.5, 3.5.6.9, verify system performance of each accumulator, core makeup tank, the passive RHR heat exchanger, and the IRWST injection and recirculation system, respectively, in accordance .with the System Level inservice Testing Program. He test methods quali'atively described in SSAR Table 3.9-17, System Level Inservice Testing Requirements, for demonstraGng the flow and heat transfer capabilities of these systems, are not sufficient

for a determination of the acceptability of the system level tests and the 10-year test interval specified.
a. ' Provide details of the test methods, and quantitative acceptance criteria along with the bases for the systern level tests of these systems.
b. Describe the programs and procedures which will ensure that the system performance will not be sevewly degraded by undetected foreign objects or debris in these systems within the surveillance imerval. ,

c.- Rese SRs specify the system level performance tests in accordance with the System Level "Insers.ce Test hogram." This is confusing because system level tests are not covered by the ASME IST program as specified in 10 CFR 50.55a. The system test frequency of each system should be specified in the TS. In addition, the so-call.:d " System Level inservice Testing" in Table 3.9-17 should be called something different so as not to confuse this program with that specified in 10 CFR 50.55a.

Response

a. Attached is a revised SSAR Table 33-17 which includes test methods and quantitative acceptance criteria. De test frequencies have been moved from this table to the Technical Specifications.

Attached are markups of the Technical Specification surveillances. The test methods contained in Table 3.9-17 need not be revised since they con *ain as much or more detail as the Standard Technical Specifications (Reference 440.667-1).

b. The Technical Specification surveillances require " system level" performance tests. The possibility

- of the system degrading between these performance tests is vety unlikely. One potential degradation mechanism for these systems is the degradation of the operability of active components (pumps and valves). The AP600 has no safety related pumps, so that is not a source of degradation. The AP600 IST program verifies the operability of safety related valves. Another possible degradation mechanism is the blocking of lines. His mechanism is very unlikely in the AP600 because the following reasons:

440.667 through 440.674, page 1 3 Westinghouse

l

-1 RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION

[

- Use of stainless steel construction eliminates corrosion products.

- Good water quality is maintamed such that there are few panicles.

- Tanks are normally closed which prevents the addition of debrir.

- Technical Specification surveillance SR 3.5.6.10 rec,uires inspection of the IRWST and containment recirculation screens every refueling.

- Combined License applicants referencing the AP600 are required to have a containment cleanliness prograin (see SSAR subsection 6.3.8.1) to prevent debris from being left in the containment following refueling or maintenance outages.

- Opening of the system for inspection or repairs is controlled by work procedures which maintain the cleanliness of the system, including the control of parts / tools to prevent them from beirig left in the system,

c. He name of this table is different from the table that identifies the ASME component level inservice testing (" Valve Inservice Test Requirements" vs " System Level Inservice Test

- Requirements"). To provide additional differentiation, the name of the system level table will be changed to " System Le 1 Operability Test Requirements". The Technical Specification references to this table will be revised to reflect this change, see attached markups.

SSAR Revision: See attached markups. .

Reference:

440.667-1: NUREG 1431.

Question 440.668 Surveillance Requirements should be added to verify the automatic response of all passive core cooling system automatic actuation valves that must change position in response to an engineered safety feature actuation signals to perform a safety function (similar to ECCS surveillance requirement in Westinghouse standard technical specifications, SR 3.5.2.5). He SR should demonstrate that each automatic valve responds and moves to the correct position on an actual or simulated actuation signal. This includes SR 3.5.2 for the CMT outlet isolation valves, SR 3.5.4 for PRHR air-operated outlet isolation valves, SR 3.5.6 fnr IRWST motor-operated recirculation isolation valves.

Response: ,-

The ESF Actuation Logic Test surveillance (SR 3.3.2.2) together with the Actuation Device Test surveillance (SR 3.3.2.7 / 8 / 9) provides overlap testing that replaces the need for a separate surveillance as required in standard Technical Specifications. This overlap is a result of the AP600 I&C design which is different than the I&C designs for which the STS were developed. During the ESF Actuation Logic Test, system level actuation signals are generated in the ESF cabinets and are sent to the component level

-logic cabinets. He automatic tester that performs the test only allows the system level signal to actuate I of the 3 redundant logic cabinet logic sets. - Since the logic sets actuate components on a 2 out of 3

. 440.667 through 440.674, page 2 W Westinghouse

~,

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION

.. =

= g basis, the components are not actuated. This approach tests the system level and the component level logic without actuating components.

The SR 3.3.2.7,3 3.2.8, and 3.3.2.9 Actuation Device Tesst will verify the automatic position change of the PCS valves using manual controls. The manual actuation signals propagate through the component logic cabinets in a manner that overlaps the path for the ESF signal. ,

The definitions of these two tests contained in the AP600 Technical Specifications will be revised to require that these tests provide overlap.

SSAR Revision: See attached Technical Specification definition changes.

Question 440.669 Surveillance Requirements should be added for the squib valves to verify continuity of explosive charge.

This includes SR 3.4.12 for ADS stage-4 squib valves, and SR 3.5.6 for IRWST i,.jection and recirculation squib valves.

Response

Technical Specification SR 3.3.2.8 is a actuation device test which is provided for squib _ valves as discussed in the bases. Table 3.3.2-1, items 25 and 26, along with the Bases SR 3.3.2.8, requires continuity testing for the squib valves used for ADS stage-4, IRWST injection and recirculation actuation.

This surveillance is performed during refueling outages in order to minimize the chance of unintended squib valve actuation.

SSAR Revision: See letter DCP/NRC0987.

Question 440.670 in the markup of 1.CO 3.5.4 on the passive residual heat removal heat exchanger, Action E requires restoration of the PRHR HX to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The Bases (E] does not provide sufficientjustification for the 8-hour completion time. The Explanation of CT/SF Value" in Table 1 (of Westinghouse's June 18,1997 letter) discussed an analysis performed to demonstrate that, if the PRHR HX fails to function during a transient or SGTR, the core can be adequately cooled using passive feed and bleed. A discussion of the specific analysis should be provided in the BASES.

In addition, there is an inconsistency between the required actions in Westinghouse's June 6, and June 18, 1997 technical specification submittals regarding insett Condition F of LCO 3.5.4. The staff believes that the required actions in the June 6,1997, submittal is correct. Please clarify the Westinghouse position 440.667 through 440.674, page 3 T Westinghouse

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION

- m -m on this change. Note that there is also a typographical error in the Condition F wording; it should read:

" Required Action and associated Completion Time for Condition E [are not met!"

Response

As presented in letter DCP/NRC0891 of 6/6/97, the PRHR HX is a unique feature of the_ AP600 and as

- such there is no directly comparable STS that is applicable. For comparison purposes, the typical plant auxiliary feedwater system (AFWS) has some similarities with the PRHR HX functionally, but has significant design differences:

-LAFWS connects to the SGs and as a result it is degraded by some of the accidents it is designed to mitigate. For example, a break of a feedwater line tends to disable one of the 3 normally provided-AFWS pumps and degrade another of the pumps. The PRHR HX is connected to the RCS and is not degraded by such accidents.

-- AFWS uses AC power for 2 pumps and steam for another pump. These pumps have to be actuated

- by the PMS. The PRHR HX uses no pumps; it only requires one of two fail open valves to open to initiate its operation. Normally these valves will be opened as a result of PMS act - ?.on. They will also open on loss of power to the air control solenoid valves or on loss of instrument air.

The AP600 is capable of mitigating DBAs without the PRHR HX by.using passive feed and bleed cooling. His feed and bleed cooling function uses the ADS for bleed and the CMTs/ accumulators /lRWST for feed. He' effectiveness of feed and bleed cooling is demonstrated in analysis performed to justify PRA success criteria. This analysis shows that for a range of events including loss of main feedwater.

SGTR,~ and small LOCAs that _ feed and bleed cooling provides adequate core cooling. - Attached is a markup of the Bases [E) to provide discussion of specific accident analysis performed without the PRHR :

HX. Rese analysis provide a high confidence that with the unavailability of the PRHR HX the core can be cooled following design bases accidents.

He required actions and completion times in the June 6,1997 submittal are correct. The Condition F wording and completion time will be revised as shown in the attached markup of this specification. Also, it has been confirmed that the other completion times and surveillance frequencies of the June 18 submittal are consistent with those of the June 6 submittal.

. SSAR Revision: See attached markups.

' 440.667 through 440.674, page 4 T Westinghouse

RESPONSES TO NRC REQUEST FOR ADDITIONAL. INFORMATION Question 440.671 In the markup LCO 3.4.12 on the automatic depressurization system Action A.! and the associated c.ompletion time are specified for the inoperability of either one now path (no stage specified), or two Dow paths consisting of one stage-1 and one of either stage-2 or stage 3 flow path. The completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> appears to be based on a Westinghouse proposed method (Letter, B. McIntyre to USNRC, NSD-NRC 96 4699, May 3,1996), which has not been approved. Since the ADS flow paths are very important for accident mitigation of the AP600 design, Westinghouse shecid propose a shorter completion for restoration of these inoperable flow paths.

Response

}I The completion time of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is NOT based on the Westinghouse proposed method of systematically specifying action times. The basis for this Action / Completion Time is consistent with the STS PORV, 3.4.11 Action B. It is also consistent with two train ECCS systems that can perform their safety function without a single failure.

The identified inoperable ADS Dow path condition of one ADS stage I and one ADS stage 3 can occur as the resuh of a single failure; the power supply for ADS stages 1 and 3 in each ADS group is from one de power supply. As a result, the unavailability of these two ADS paths is a credible single failure and is considered in SSAR chapter 15 LOCA analysis. Also, the limiting single failure in the chapter 15 LOCA analysis is one ADS stage 4 valve failure because its now capability is much greater than one ADS stage 1 and one stage 3 path (refer to SSAR subsection 15.6.5.4B.I.0). 'Ihe response to RAI 440.662 -

shows that a DVI LOCA with one ADS stage I and one stage 3 path failed has adequate core cooling, essentially the same as shown in the SSAR (subsection 15.6. 5.4B.3.3) with the failure of one ADS stage

- 4 valve.

Since ADS stage 2 and stage 3 are identical valves and have the same depressurization capacity, the technical specification is written to allow either to be inoperable along with one ADS stage 1.

SSAR Revision: None Question 440.672 LCO 3.5.1 on accumulators specifies a completion time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the. Required Action B.1 to restore the inoperable accumulator. Since this is longer than that specified in the standard TS (NUREG-1431),

the basis for this longer completion time should be included in the TS BASES. This also applies to the LCOs for other ECCS systems, such as LCO 3.5.2 through 3.5.8 for Core Makeup Tanks. PRHR heat exchanger, and IRWST, respectively.

440.667 through 440.674, page 5 3 Westliighouse

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION i

Response

See the response to RAI 440.670 for justification of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the PRHR HX. The IRWST only uses 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for boron concentration and temperature deviations which is consistent with the STS; it uses I hour for complete system unavailability (see Action 3.5.6 D). The following justifies 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the accumulator and the CMT.

Accumulator:

As presented in letter DCP/NRC0891 of 6/6/97, the use of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> instead of I hour (STS LCO 3.5.1) is justified based on AP600 design differences that provide significantly greater margin in LOCA perfcrmance than current plants with one accumulator unavailable.

' In current 2 loop plants, the accumulators are connected to the cold legs such that a large LOCA can result in the spilling of one accumulator. If the other accumulator is unavailable prior to a large LOCA, no accumula. ors would be available to mitigate the large LOCA. As a result, the Technical Specifications for such a design should not allow operation for a significant length of time with an accumulator unavailable.

The AP600 accumulators are connected directly to the reactor vessel thro' ugh small bore DVI lines rather than to the RCS large bore cold legs. As a result, a large LOCA does not cause an accumulator to spill and be lost. If one accumulator is unavailable prior to a large LOCA, the other accumulator is still available to inject directly into the reactor vessel via the DVI lines. Such a design could operate for a significant length of time with one accumulator unavailable, provided it is demonstrated that one accumulator provides sufficient flow to mitigate the range of design basis LOCAs.

Such analysis has been performed in order to justify the PRA success criteria. This analysis shows that with one unavailable and the other accumulator injecting for a large LOCA that the core is adequately cooled. It also shows that with no accumulators injecting for small LOCAs that the core is adequately cooled. This analysis was performed with the same codes used for the SSAR analysis.

LCO 3.5.1 Action B.1 will be revised, as shown in the attached markup, to provide discussion of specific accident analysis puformed with one accumulator unavailable. These analysis provide a high confidence that with the unavailability of one accumulator the core can be cooled following design bases accidents.

ChfTS As presented in letter DCP/NRC0891 of 6/6/97, the CMTs are a unique feature of the AP600 and as such there is no directly comparable STS that is applicable. The HHSI pumps have some similarities with the CMTs functionally, however it has significant differences in design including:

440.667 through 440.674, page 6 W WestinEhouse

RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION

- The HHSI pumps take suction from one tank (RWST) and each HHSI pump injects into multiple locations. The HHS1 pumps are arranged so that one can be lost to a single failure and the other pump can deliver sufficient Si flow even when it is spilling flow out one branch line (caused by LOCA). The HHSI pumps are required to continue to operate indennitely which requires them to switch their suction from the RWST to the containment sump. The AP600 Ch1Ts contain their own supply of water. Each CMT only injects to one RCS locatio' ah Chit has redundant actuation valves. The CMTs are designed such that one Ch1T can be lost opill) to a LOCA ef the DVI line and the other CMT can tolerate a single failure and still deliver sufGeient SI flow.

The AP600 can tolerate the complete loss of the CMTs and mitigate DBAs. In such cases the accumulators and IRWST injection provide sufficient injection to cool the core. The effectiveness of accumulator and IRWST injection (without CMTs) is demonstrated in analysis performed to justify PRA success criteria. This analysis shows that for a range of events including SGTR and small LOCAs that i

these other injection supplies provide adequate core cooling. Attached is a markup ofLCO 3.5.2 Action E.1 to provide discussion of specific accident analysis performed without the CMTs. These analysis provide a high confidence that with the unavailability of the CMTs the core can be cooled following design bases accidents.

SSAR Revision: See attached markups.

Question 440.673 Condition C of LCOs 3.5.6, 3.5.7, and 3.5.8, on the IRWST include a " deviation of water volume not within its limit" as the inoperability of the IRWST, and specifies a completion time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the restoration of the IRWST. The justification provided in the B ASES to allow 8-hour completion time is not acceptable since the exp'anation is related to " minor" deviation in the water volume whereas the LCO Condition C does not specify " minor" deviation in water volume. Herefore, the " water volume" should be deleted from Condition C so that if the IRWST water volume is not within the acceptable limit, Required Action should be taken per Condition E.

Response

Condition C of LCbs 3.5.6,3.5.7, and 3.5.8 has been marked up to limit the IRWST volume deviation to 3% (volume <100% and >97% of the LCO limit).

The LCO 3.5.6,3.5.7, and 3.5.8, Condition C Bases have also been marked to justifyan 8-hour restoration time for a minor deviation in IRWST water volume of less than 3%. This justification is based on thermal / hydraulic analysis performed to justify the PRA success criteria (WCAP-14800). This analysis demonstrates that the post-accident, long term cooling supported by the IRWST is accomplished with a ,

4,000 cubic foot reduction in the water volume in containment. His analyzed cooling water reduction envelopes the Condition C IRWST vo!ume deviation of less than 3%. Therefore, the 8-hour Completion 440.667 through 440.674, page 7 Vj Westinghouse

O RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION

.tiu..m Time is acceptable based on the ability of the IRWST to perform its safety function as analuzed in WCAP 14800.

SSAR Revision: See attached markups.

Question 440.674 IRWST surveillance 3.5.6.3 requires ver'fication of the IRWST boron concentration within the acceptable limit once per 31 days, which is inconsistent with 7 days specified in STS. The justification provided in

" Explanation of CT/SF Value" is not acceptable. His SR should be revised with a frequency of 7 days.

Response

As presented in letter DCP/NRC0891 of 6/6/97, there are design differences between the AP6001RWST and the RWST design used for the STS, These design differences include location of the tank inside containment and the use of a larger tank (20 to 60% larger). Location of the tank inside containment makes it less likely that water will be added to the tank, diluting the boron solution.

Two ways of changing the IRWST boron concentration have been identif'ied, one is the addition of dilute water to the IRWST and the other is recirculation of the IRWST with a dilute water volume. The only IRWST recirculation paths are with the RNS and the SFS. he RNS can not change the IRWST boron concentration by recirculation because it is not connected to another tank and the volume of its pipes, pumps, and HX is insignificant with respect to the IRWST. The spent fuel pool is a large volume however its boron concentration is the same as the IRWST, so recirculation with it can not reduce the IRWST below its limits.

It is possible to dilute the IRWST boron concentration by adding dilute water to the tank. Adding enough dilute water to increase the IRWST volume by 3 % (- 15,000 gal) would reduce the IRWST boron concentration by about 75 ppm. The IRWST boron concentration limits are 2600 ppm to 2900 ppm.

Such a small change in boron concentration (- 3 %) is insignificant with respect to the shutdown margins provided. SR 3.5.6.3 requires the IRWST boron concentration to be verified whenever its water volume has been increased by 15,000 gal. The STS( SR 3.5.4.3) do not have such a reouirement.

Since the only credible way of significantly reducing the IRWST boron concentration is to add dilute 4 water to the IRWST, monitoring the IRWST water volume provides an effective means of preventing a significant boron reduction. It is proposed that the IRWST water volume surveillance be made every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in order to reduce the possibility of the boron concentration being reduced below its limits; the STS (SR 3.5.4.2) only require the IRWST level be surveyed every 7 days. Increasing the IRWST volume by 15,000 gal in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> would require an average flow of 10 gpm which will be readily detectable by other means such as RCS leakage or CVS makeup measurements.

440.667 through 440.674, page 8 3 Westinghouse

. RESPONSES TO NRC REQUEST FOR ADDITIONAL INFORMATION d am

=

m Providing surveillance of the IRWST volume every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> together with the requirement to verify the boron concentration after a volume increase of 3% provides high confidence that the IRWST boron concentration will not be reduced to below its limits. These additional surveillances justify the use of a 31 day suneillance of the IRWST boron concentration.

SSAR Revision: None.

O 440.667 throt:gh 440.674, page 9 l

- - _ _ _

Retrieved from "https://kanterella.com/w/index.php?title=NSD-NRC-97-5278,_Forwards_Responses_to_NRC_Reactor_Sys_Branch_RAIs_on_AP600_Ts,Including_Resultant_Ssar/Ts_Markups&oldid=3339539"