ML17313A601

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Provides Response to Request for Addl Info Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions. with twenty-nine Oversize Drawings
ML17313A601
Person / Time
Site: Palo Verde  
Issue date: 09/16/1998
From: James M. Levine
ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17313A602 List:
References
102-04183-JML-S, 102-4183-JML-S, GL-96-06, GL-96-6, NUDOCS 9809230249
Download: ML17313A601 (57)


Text

'ATEGORY( 1 REGULATORY INFORMATION DISTRIB ION SYSTEM (RIDS)

ACCESSION NBR:9809230249 DOC.DATE: 98/09/16 NOTARIZED: NO DOCKET I FACIL:STN-50-528 Palo Verde Nuclear "Station, Unit 1, Arizona Publi ~05000528 STN-50-529 Palo'Verde Nuclear Station, Unit 2, Arizona Publi 05000529 STN-50-530 Palo Verde Nuclear Station, Unit 3, Arizona Publi 05000530 AUTH.NAME AUTHOR AFFILIATION LEVINE,J.M.

Arizona Public Service Co.

(formerly Arizona Nuclear Power RECIP.NAME RECIPIENT AFFILIATION Records Management Branch (Document Control Des C

SUBJECT:

Provides response to regnest for addi info re GL 96-06, "Assurance of Equipment Operability

& Containment Integrity During Design Basis Accident Conditions." With twenty-nine oversize drawings.

D1STRTBUT10N CODE:

A072D COPTES RECE1'VED: LTR J ENCL J STBE:

TITLE: GL 96-06, "Assurance of Equip Oprblty 6 Contain.Integ.

during Design A

NOTES: STANDARDIZED PLANT Standardized plant.'tandardizedplant.

05000528 05000528

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RECIPIENT ID CODE/NAME NRR/WETZEL,B.

FIELDS M COPIES LTTR ENCL 1

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1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE. TO HAVE YOUR NAME OR ORGANIZATION REMOVED FROM DISTRIBUTION LISTS OR REDUCE THE NUMBER OF COPIES RECEIVED BY YOU OR YOUR ORGANIZATION, CONTACT THE DOCUMENT CONTROL DESK (DCD)

ON EXTENSION 415-2083 TOTAL NUMBER OF COPIES REQUIRED:

LTTR 9

ENCL 9

Palo Verde Nuclear Generating Station James M. Levine Senior Vice President Nuclear TEL (602)393-5300 FAX(602)393-6077 Mail Station 7602 P.O. Box 52034 Phoenix, AZ 85072-2034 102-04183-JML/SAB/RMW September 16, 1998 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station P1-37 Washington, DC 20555

Dear Sirs:

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3 Docket Nos. STN 50-528/529/530

Response

to Request for Additional Information Regarding Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity during Design Basis Accident Conditions".

. The enclosure to this letter provides the additional information regarding Arizona Public Service Company's (APS) response to Generic Letter (GL) 96-06, "Assurance of Equipment Operability and Containment Integrity during Design Basis.Accident Conditions" for the Palo Verde Nuclear Generating Station, Unit Nos. 1, 2 and 3, that was requested in your letter to APS dated June 24, 1998.

The enclosure to this letter commits APS to revise the appropriate station emergency operating procedures to provide guidance to the operations staff that would prohibit restoration of the Nuclear Cooling and Normal Chilled Water systems after certain plant conditions occurred, without first performing a fill and vent of these systems.

The revisions to these procedures will be completed during the next scheduled update for these procedures, which is currently scheduled for completion in December 1999.

Please contact Mr. Scott Bauer at (602) 393-5978 if you have any questions or would like additional information regarding this matter.

Sincerely, JML/SAB/RMW/IIh Enclosure O

( so~:)

cc:

E. W. Merschoff M. B. Fields J. H. Moorman 9809230249

'F80916 PDR ADOCK 05000528 PDR

0

ENCLOSURE Response to Request for Additional Information Regarding Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions", for the Palo Verde Nuclear Generating Station.

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~RRR R The licensee's response indicated that the containment cooling water systems at Palo Verde are not safety-related and are not susceptible to waterhammer or two-phase flow conditions.

In order to assess the licensee's resolution of these issues, the following additional information is required:

Note: The following questions are applicable to cooling water systems associated with the in-containment environmental heat removal function.

NRC Re uest 1 Implementing measures to assure that waterhammer and two-phase flowwill not occur, such as controlling post-accident operation of systems that could be

affected, is an acceptable approach for addressing these concerns.
However, all scenarios must be considered to assure that the vulnerability to waterhammer and two-phase flow has been eliminated.

Confirm that all scenarios have been considered, including those where the affected containment penetrations are not isolated (if this is a possibility), such that the measures that have been established are adequate to prevent the occurrence of waterhammer and two-phase flow during (and following) all postulated accident scenarios.

Identify the "worst-case" conditions that could occur within the licensing basis of the plant, such as event possibilities, system configurations, parameters, and component failures, in order to adequately assess the waterhammer and two-phase flow concerns.

As an example, the 120-day response did not address the scenario where a thermal relief valve has lifted, possibly reducing the pressure of the affected cooling water system to below saturation.

~SPS R Introduction References 10 and 11 provided APS'esponse to Generic Letter 96-06, "Assurance of Equipment Operability and Containment Integrity during Design Basis Accident Conditions".

These references indicated that the essential and non-essential containment cooling water systems at PVNGS were not vulnerable to waterhammer or two-phase flow conditions that may cause damage to these systems during design basis accident conditions.

As stated in these references, the safety-related Containment Spray (CS) system, which is an open spray system, is credited for maintaining containment pressure and temperature within the appropriate design limits should a design basis accident occur.

This system is properly analyzed for the post-design basis event containment environment, inclusive of potential waterhammer or Page 1

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two-phase flow conditions.

Therefore, the safety-related function of this system is not impacted due to post-accident containment environmental conditions.

During normal operating conditions, the containment environment heat removal function is provided by the non-safety-related, closed-loop, normal Chilled Water (WC) and Nuclear Cooling Water (NC) systems.

Although these systems are not required during post-accident conditions, restoring these systems to operation may facilitate post-accident recovery operations.

Therefore, an evaluation of these systems was performed to demonstrate that these systems would not be susceptible to waterhammer or two-phase flow conditions during the predicted accident scenarios, or to provide guidance to the operations staff such that these systems may be safely returned to service during post-accident recovery operations in a manner that would preclude waterhammer or two-phase flowconditions.

APS has limited the discussion provided in response to your requests to these two systems since these systems are the only closed-loop containment cooling water systems that are associated with the in-containment environmental heat removal function.

Res onse to Re uest 1 The setpoints and design characteristics of the relief valves associated with the WC and NC systems are such that subcooled fluid conditions will be maintained within their respective system when the in-containment piping is exposed to those conditions that could potentially result in lifting of a relief valve.

The setpoints and design characteristics for the relief valves are described in the responses to NRC Requests 3

and 4.

In accordance with the PVNGS UFSAR failure analyses methodologies, failure of a relief valve is not considered to be a credible failure as further explained in the response to NRC Request 2.

The WC containment isolation valves, valves 13JWCAUV0061, 13JWCAUV0062, and 13JWCBUV0063, automatically close upon receipt of a CIAS. The CIAS is generated at a nominal containment pressure of 3 psig, or a nominal pressurizer pressure of 1837 psia (Reference 1, Table 7.3-11A).

The NC containment isolation valves, valves 13JNCBUV0401, 13JNCAUV0402, and 13JNCAUV0403, automatically close upon receipt of a CSAS.

The CSAS is generated at a nominal containment pressure of 8.5 psig (Reference 1, Table 7.3-11A).

Although these systems are not credited for the mitigation of any design basis accident (Reference 1,

Sections 9.2.2.2.1 and 9.2.9.1.1.1),

recovery from design basis accidents may be operationally enhanced if these systems are capable of being returned to service such that cooling water is provided to the in-containment heat loads.

The current PVNGS emergency operating functional recovery procedure (Reference

2) provides guidance for returning these systems to service.

The current guidance specifies that the following conditions must be met prior to restoring cooling water flowto the in-containment heat loads:

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1) a Recirculation Actuation Signal (RAS) has not occurred, and
2) containment pressure is less than 8.5 psig.

Three design basis event scenarios which could result in thermally induced waterhammer or two-phase flowconditions in the containment cooling water systems at PVNGS are: Large Break Loss of Coolant Accident (LBLOCA), Main Steam Line Break (MSLB), and Small Break Loss of Coolant Accident (SBLOCA). These accident scenarios will result in one of three post-accident system configurations for the NC and WC systems.

The expected plant response, with respect to the initiation of CIAS, CSAS and RAS, associated with these accident scenarios is briefly described below.

Immediately following this description, the three potential post-accident system configurations, and resultant vulnerability to waterhammer or two-phase flow conditions, are described.

LBLOCARes onse The pressure-temperature profiles for LBLOCA scenarios are presented in Reference 1, Figures 6.2.1-1 through 6.2.1-5.

In each case, containment pressure rises above the CIAS and CSAS setpoints within the initial seconds of the event.

In four of the five

cases, a

RAS occurs before containment pressure is reduced below 8.5 psig.

Therefore, the WC and NC system containment isolation valves are closed from the start of these events.

Since a RAS does occur, the current PVNGS emergency operating procedures prohibit the restoration of the WC and NC cooling water flow to in-containment heat loads.

Therefore, the containment isolation valves for these systems remain closed and are not re-opened during post-accident recovery operations.

In these accident scenarios, the response of the WC and NC systems will be as described in References 10, 11 and 12.

The LBLOCA analysis presented in Reference 1, Figure 6.2.1-1 is not extended to evaluate long-term plant conditions because the resultant energy release to containment is low and this analysis is bounded by other analyses (Reference 3).

Therefore, in this LBLOCA scenario, it is possible that the containment pressure will decrease below 8.5 psig without a RAS occurring first.

In accordance with the current PVNGS emergency operating procedures, the WC and NC systems may be restored, i.e., the containment isolation valves may be

reopened, to enhance operational recovery from this accident scenario.

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MSLB Res onse The pressure-temperature profiles for MSLB scenarios at various power levels are presented in Reference 1, Figure 6.2.1-6.

In each case, containment pressure rises above the CIAS and CSAS setpoints within the initial seconds of the event.

However, the long-term post-accident containment pressure is reduced without the initiation of a RAS. Therefore, the WC and NC containment isolation valves, which would be isolated at the beginning of a MSLB, could then be subsequently re-opened during post-accident recovery operations in accordance with the current PVNGS emergency operating procedures.

SBLOCA Res onse In addition to the MSLB and LBLOCA analyses contained in Reference 1, Section 6.2.1, Emergency Core Cooling performance evaluations for various LBLOCA break

sizes, as well as SBLOCA scenarios, are described in Reference 1, Section 6.3.3.

Reduced LOCA break sizes have the effect of decreasing containment heatup and pressurization rates.

Therefore, for small SBLOCA scenarios, the CIAS and CSAS actuations will be delayed from the onset of the accident, during which time the WC and NC containment isolation valves will remain open.

Depending on the break size, a RAS may, or may

not, occur.

Additionally, as the magnitude of pressure and temperature effects decrease with reduced break sizes, the potential for post-accident re-opening of the WC and NC containment isolation valves in accordance with the current PVNGS emergency operating procedures, increases.

The design basis event scenarios described above will result in one of the following three post-accident system configurations for the in-containment WC and NC systems:

1. The WC and NC systems'ontainment isolation valves are automatically closed at the start of the event in response to a CIAS and CSAS and are not re-opened during post-accident recovery operations (LBLOCAscenarios).
2. The WC and NC systems'ontainment isolation valves are automatically closed at the start of the event in response to a CIAS and CSAS and are subsequently re-opened during post-accident recovery operations in accordance with the current PVNGS emergency operating procedures (MSLB, some LB and SB LOCA scenarios potentially affected).
3. The WC and NC systems containment isolation valves remain open for some time after the initiation of the event (small SBLOCA scenarios).

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Each of these post-accident system configurations is described in the following paragraphs.

WC and NC Containment Isolation Valves Remain Closed throu hout the Accident In these accident scenarios, water within the WC and NC systems would be trapped within the in-containment piping and would thermally expand due to the accident conditions.

As described above, the WC and NC systems thermal relief valves have a combined relieving rate that has been determined to be adequate to discharge the amount of water required such that over-pressurization of the affected in-containment piping sections is not expected to occur during accident conditions.

In addition, the relief valve setpoints are higher than the saturation pressure of the post-accident heated water in the system (Reference 5).

Thus, the formation of steam voids is prevented, which precludes the development of waterhammer or two-phase flow conditions.

WC and NC Containment Isolation Valves Close at Event Initiation and are Re-0 ened durin Post-Accident Recove 0 erations As described in Reference 10, voiding within the WC and NC systems does not occur under this condition due the fact that the fluid within these systems remains pressurized above the applicable peak saturation pressures.

The in-containment piping of these systems remain pressurized for two reasons:

1. The capacities of the relief valves are sufficient to ensure that the in-containment piping is not damaged due to over-pressurization of the piping.

In addition, the setpoints of the relief valves are such that the fluid within these systems is maintained in the subcooled state (Reference 5).

2.

Each system is verified to be operating, i.e., pump(s) running, prior to opening the associated containment isolation valves.

In addition, the sequence for opening the containment isolation valves is specified such that the normal system operating pressure is maintained within the in-containment system piping (Reference 2,

Appendix 17, and Reference 8). At the time the isolation valves are re-opened, the system operating pressures are above the post-accident saturation pressure corresponding to the predicted containment temperature.

Therefore, voiding is not expected to occur under this configuration and waterhammer or two-phase flow is precluded.

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WC and NC Containment Isolation Valves Remain 0 en for Some Time after Event Initiation Two potential scenarios are associated with this system configuration.

These scenarios are:

1) either, or both, of the systems are in operation, and an actuation signal (CIAS or CSAS) may, or may not, be received that results in isolating the in-containment piping of the affected system, or 2) either, or both, of the systems are not in operation when the event is initiated, or stop operation prior to the affected system's containment isolation valves being closed.

When the design basis event occurs during the first scenario, the system is operating and water is not trapped within the in-containment piping.

The WC and NC systems continue to operate throughout the event.

As the event progresses, containment temperature and pressure begin to increase, heating the in-containment piping of these systems.

The normal operating pressures associated with these systems are such that the fluid within the system is maintained in a subcooled state until the system is isolated automatically by a CIAS or CSAS actuation signal.

The WC containment isolation valves may close on a CIAS, followed by the possible closure of the NC containment isolation valves on a CSAS, if the break size is such that containment pressure increases to the signal actuation setpoint.

These actuation signals will isolate the in-containment portion of the system piping from the rest of the system.

The containment environment will be at saturation conditions for a nominal saturation pressure of 8.5 psig when both systems have been isolated.

Since the in-containment piping of each system was isolated at the normal system operating pressure associated with the system, voiding is not expected to occur and the system response is expected to be as that which is described in References 10, 11 and 12.

Ifthe size of the break is such that containment pressure does not increase to the actuation signals'etpoint, then the normal operating pressure of the system will ensure that the fluid within the in-containment piping of the system remains in a sub-cooled state.

If either of these systems have been isolated from their respective containment heat loads during the scenario described above, the current PVNGS emergency operating procedures would direct the operator to restore cooling water flow, if desired, to the in-containment heat loads of the affected system provided that a RAS has not occurred.

In accordance with the current PVNGS emergency operating procedures, the WC and NC systems are verified to be in operation, i.e. pumps running, before the containment isolation valves are re-opened during post-accident recovery operations.

System malfunctions that may occur prior to re-opening the containment isolation valves that would preclude system availability would result in the affected system's containment isolation valves remaining closed (Reference 2, Appendix 17, and Reference 8).

Therefore, two-phase flow or waterhammer is not expected to occur when the affected system is restored.

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When the design basis event occurs during the second scenario, the system is secured and water is not isolated within the in-containment piping of the system until a CIAS (for WC) or CSAS (for NC) is generated.

The water within the in-containment piping of the affected system would expand throughout the closed loop system and into the system's surge tank until the CIAS and/or CSAS actuation signals are generated.

However, the static, non-operating system pressures associated with the WC and NC systems may not be sufficient to maintain subcooling of the fluid within the in-containment piping of the affected system. The postulated scenarios where voiding may be expected to occur within these systems are limited to small, SBLOCA scenarios and small, high energy line break in containment scenarios that result in, the following conditions occurring simultaneously:

1.

Containment environment temperature has exceeded 212 'F,

2. The containment isolation valves for the affected system (WC or NC) are open, and
3. The affected system is not in operation.

However, the potential for waterhammer or two-phase flow conditions exists only if the affected system were restarted during, or following, these three plant conditions occurring.

The WC and NC pumps do not receive a post-accident automatic start signaI.

Therefore, the only method for restoring these systems to operation is through manual operator actions.

Although the current PVNGS emergency operating procedures permit restoration of these systems during post-accident recovery operations when certain plant conditions are met, these procedures do not prohibit restoration of these systems when the above three plant conditions are experienced.

The appropriate PVNGS emergency operating procedures will be revised to provide guidance to the operations staff that would prohibit restoration of the WC and NC systems after the above plant conditions occurred, without first performing a fill and vent of these systems.

The revisions to these procedures will be completed during the next scheduled update for these procedures, which is currently scheduled for completion in December 1999.

These procedure revisions will assure that waterhammer or two-phase flow conditions that may occur as a result of these limited accident scenarios are eliminated prior to restoring the WC or NC system to operation.

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NRCRe uest2 Confirm that the waterhammer and two-phase flowanalyses included a complete failure modes and effects analysis (FMEA) for all components (including electrical and pneumatic failures) that could impact performance of the cooling water system and confirm that the FMEA is documented and available for review, or explain why a complete and fully documented FMEA was not performed.

A~PS R A FMEA was not performed for the WC and NC systems as the systems were determined not to be susceptible, either by system design or procedural control, to waterhammer or two-phase flow conditions that may result from a containment thermal transient event.

Therefore, adverse system effects resulting from waterhammer or two-phase flow are not expected to occur within these systems.

As described in the response to NRC Request 1, system design and adequate station procedural guidance prevent waterhammer or two-phase flow conditions from occurring when the post-accident system configuration is such that the containment isolation valves for the affected system are automatically closed at the start of the event in response to a CIAS and CSAS.

The WC and NC systems are procedurally required to be in operation, such that the system piping outside of containment is at normal operating pressure, before the containment isolation valves are re-opened during post-accident recovery operations.

The capacities and blowdown setpoints of the installed thermal relief valves on the WC and NC in-containment piping are relied upon to preclude the development of waterhammer or two-phase flow conditions within these

systems, after the affected system's in-containment piping has been isolated.

The mechanical aspects associated with these valves are discussed further in the responses to Requests 3 and 4 provided below.

In accordance with failure analyses provided in Reference 1, failure of a relief valve is not considered to be credible.

Per Sections 5.4.7.3 and 6.3.2.5.4 of the Combustion Engineering Standard Safety Analysis Report (CESSAR) (incorporated by reference into Sections 5.4.7 and 6.3.2.5 of Reference 1), the FMEA performed for the shutdown cooling and safety injection systems were based on the assumption that relief and check valve failures are not considered to be credible.

Additionally, CESSAR Section 5.2.2.10.1.2 (incorporated by reference into Reference 1, Section 5.2) states that the shutdown cooling system relief valves meet the single failure criteria required by CESSAR Section 5.4.7.1.2.

The CESSAR notes that the relief valves are independent of a loss of offsite power, are self-actuating, spring-loaded liquid relief valves, and they do not require control circuitry. These features also apply to the relief valves of the WC and NC systems.

Therefore, it is appropriate, and consistent with Page 8

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Reference 1, to apply these same criteria to the WC and NC relief valves.

Based on this information, failure of the WC or NC relief valves is not considered credible.'he post-accident system configuration where the WC and/or NC containment isolation valves remain open for a period of time following the initiation of the event with the affected system's pump idle may result in voiding of that system.

As stated previously, APS will revise the appropriate station emergency operating procedures to provide guidance to the operations staff that would prohibit restoration of the WC and NC systems after certain plant conditions occurred, without first performing a filland vent of these systems NRC Re uest 3 Discuss specific system parameter requirements that must be maintained in order to assure that waterhammer and two-phase flow will not occur (e.g.,

expansion tank level, temperature, and pressure requirements),

and state the minimum margin to boiling that exists, including consideration of measurement and analytical uncertainties.

Explain why it would not be appropriate to establish technical specification requirements for these parameters, and describe and justify reliance on any non-safety related instrumentation and controls for assuring that waterhammer and two-phase flowwillnot occur.

APS Res onse The WC and NC systems are non-safety-related systems that are not required to function during accident conditions or for post-accident recovery operations.

As such, these systems do not meet the requirements of 10 CFR 50.36 and should not be included in the Technical Specifications.

However, recovery from design basis accidents may be operationally enhanced if these systems are functional during post-accident recovery operations.

Therefore, the scenarios evaluated above were performed for the purpose of recognizing the cases that could potentially result in the conditions being established for waterhammer or two-phase flow to occur.

For those cases that could potentially lead to waterhammer or two-phase flow, procedural controls have been, or will be, established to return the system to service in a manner that prevents waterhammer or two-phase flow.

As previously discussed, installed thermal relief valves are relied upon to preclude the development of waterhammer or two-phase flow conditions within the WC and NC systems.

The uncertainty associated with the relief valve set pressure is +/- 3 %, with a blowdown of 10%

of set pressure (Reference 5).

Incorporating set pressure uncertainties results in a lowest reseat pressure of 131 psig for the WC relief valves and 96 psig for the NC relief valves.

Reference 5 conservatively calculated the WC Page 9

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and NC relief valve capacities utilizing a post-accident containment vapor temperature of 300'F (Reference 6), which has a corresponding saturation pressure of 52.8 psig.

Therefore, the margin to boiling is 78.2 psig for the WC system and 43.2 psig for the NC system.

Smaller break size LOCA scenarios were discussed in the response to Request 1

wherein flow through the in-containment piping of the WC and NC systems would continue from the initiation of the event until the containment isolation valves are actuated closed.

As previously noted, both systems would have been isolated once containment reached saturation conditions corresponding to the nominal CSAS actuation pressure of 8.5 psig.

Note that the WC system isolates on a CIAS, which is initiated at a nominal containment pressure of 3 psig; however, use of the CSAS limit is bounding and, therefore, conservative.

The analytical limit for the CSAS actuation is 10 psig (Reference 9). Therefore, the maximum uncertainty for the margin to boiling for this scenario is given by the containment saturation conditions associated with 10 psig (238'F).

Per Reference 8, the minimum WC in-containment operating pressure is 50 psig.

The minimum NC in-containment operating pressure is 29 psig (Reference 4).

Therefore, the minimum margin to boiling is 40 psig for the WC system and 19 psig for the NC system.

NQR instrumentation and controls are not relied upon to preclude waterhammer or two-phase flow conditions in the WC and NC systems.

PVNGS does not consider it appropriate to establish technical specification requirements for the NC and WC relief

valves, as failure of these valves is not considered credible.

This assumption is consistent with the treatment of similar relief valves (Reference 1).

NRC Re uest 4 Explain and justify all uses of "engineering judgment" that are credited in the waterhammer and two-phase flowevaluations.

APS Res onse The following "engineering judgments" were used for these evaluations:

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The adequacy of the installed WC and NC relief valves was evaluated in Reference 5 utilizing a post-accident containment vapor temperature of 300 'F.

This temperature was used based on Reference 6, which determined that the limiting case LOCA containment maximum temperature is 293.5 'F.

Although the peak vapor temperatures for a MSLB exceed 300'F, the effect on the cold fluid isolated within the in-containment piping of the WC and NC systems is not as severe due to the short duration of the superheat and the slow thermal response of the piping Page 10

(Reference 5).

Utilization of a 300'F constant containment vapor temperature was chosen to simplify the solution for the heatup rates and subsequent water expansion within the in-containment piping portions of the WC and NC systems.

Additionally, the rate of heatup and expansion of the water isolated in the containment coolers (containment normal air coolers

[WC] and Control Element Drive Mechanism

[CEDM] coolers

[NC]) was assumed to be the same as the surrounding environment.

Actual maximum system water temperatures and heatup rates would be less than those associated with the containment vapor due to the lag in thermal response associated with the WC and NC piping and components.

Therefore, utilization of the maximum containment vapor temperature of 300'F is conservative and will over-predict the rate of heatup and resultant expansion of water trapped within the in-containment piping of these systems.

A constant heat transfer coefficient of 250 BTU/hr-ft'-'F on the external surface was assumed for NC piping submerged in post-LOCA sump water (References 5 and 7).

This constant heat transfer coefficient conservatively bounds the condensing heat transfer coefficient determined in Reference 6 and reported in Reference 1, Figure 6.2.1-7, and would be representative of the heat transfer coefficient assuming the entire pipe was submerged in the post-LOCA sump fluid.

o A review of Reference 1, Figure 6.2.1-7 indicated that the condensing heat transfer coefficient remained below approximately 130 BTU/hr-ft'-'F, after the initial 30 seconds of a LOCA. Hence, for the NC piping exposed to post-LOCA containment air during the initial period of 0-60 seconds, a constant heat transfer coefficient of 200 BTU/hr-ft'-'F on the pipe external surface was assumed.

For time periods greater than 60 seconds following a LOCA, a constant heat transfer coefficient of 130 BTU/hr-ft'-'Fwas assumed (Reference 5).

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A heat transfer coefficient of 0.263 BTU/hr-ft'-'F was utilized for the insulated WC piping. This assumption is based on one inch of fiberglass insulation with a thermal conductivity of 0.022 BTU/hr-ft-'F.

Similarly, a heat transfer coefficient of 4.58 BTU/hr-ft'-'F was utilized for submerged insulated WC piping, based on one inch of water saturated fiberglass insulation with a thermal conductivity of 0.4 BTU/hr-ft-'F (Reference 5).

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A constant, internal, convective heat transfer coefficient of 125 BTU/hr-ft'-'F is assumed between the piping and the isolated water within the pipe per Reference 7.

Additionally, it is assumed that all fluid internal to the pipe is uniformly mixed and increases in temperature homogeneously.

e The evaluation of the NC system relied upon the blowdown pressure reseat values and relief capacities associated with relief valves 13JNCNPSV0414, 415, 416, 417, 480 and 499.

However, the NC system is also provided with two large relief valves, Page 11

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valves 13JNCEPSV0614 and 13JNCEPSV0615, that are designed to protect the NC containment penetration and isolation devices from being over-pressurized in the event of an intersystem LOCA via the reactor coolant pump seal injection high pressure coolers.

These two relief valves have a set pressure of 135 psig +I-3%,

with a blowdown of 50% of set pressure, such that the minimum set pressure is 131 psig.

An additional 2.3 psig reduction is assumed due to the elevation difference between these two relief valves and relief valves 13JNCNPSV0414, 415, 416, 417, 480 and 499.

Therefore, the minimum set pressure for valves 13JNCEPSV0614 and 13JNCEPSV0615 is 128.7 psig.

Note that the minimum reset pressure (64.4 psig) for these two valves is above the post-accident containment vapor pressure of 52.8 psig.

Therefore, these valves will reseat such that the NC system water remains subcooled.

Note also that the minimum set pressure of these two valves, 128.7 psig, is above the maximum expected accumulation of 124.6 psig for valves 13JNCNPSV0414, 415, 416, 417, 480, and 499.

Therefore, these two valves, valves 13JNCEPSV0614 and 13JNCEPSV0615, are not expected to open due to the fact that valves 13JNCNPSV0414, 415, 416, 417, 480, and 499 have sufficient relief capacity to prevent NC system pressurization above the expected maximum accumulation pressure (Reference 5).

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The NC piping located within the containment building contains approximately 88%

of its volume in 6, 8 and 10-inch diameter header piping. As 6-inch diameter piping has the smallest thermal mass to surface area ratio of any of these pipe sizes, it was selected as the basis of the heatup analysis.

Any non-conservatism introduced by ignoring the smaller thermal mass to surface area ratios of the portions of these systems less than 6 inches in diameter are judged to be more than counter balanced by the conservative temperature profiles and heat transfer coefficients that are being assumed (Reference 5).

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NRC Re uest 5 Provide a

simplified diagram of the affected

systems, showing major components, active components, relative elevations, lengths of piping runs, and the location of any orifices and flowrestrictions.

APS Res onse Piping and Instrument Diagrams (P8 IDs) of the PVNGS Unit 1 WC and NC systems are provided, showing the in-containment piping as well as the basic flow loop outside of containment.

Piping Isometric Drawings are also provided.

Only minor variations exist between the three PVNGS units in the WC and NC systems.

Thus, the Unit 1 drawings are representative of all three PVNGS units.

l~

The following drawings are provided:

01-M-NCP-001 01-M-NCP-003 01-M-WCP-001 01-M-EWP-001 01-P-NCF-1 15 01-P-NCF-1 16 01-P-NCF-1 17 01-P-NCF-160 01-P-NCF-1 64 01-P-NCF-202 01-P-NCF-207 01-P-NCF-209 01-P-WCF-123 01-P-WCF-'I 24 01-P-WCF-1 66 01-P-WCF-1 67 01-P-WCF-201 01-P-WCF-202 01-P-WCF-205 01-P-WCF-206 01-P-ZYA-010 13-P-ZYA-013 13-P-ZYA-015 Rev. 5 Rev. 8 Rev. 14 Rev. 24 Rev.

1 Rev.

1 Rev.

1 Rev.

1 Rev. 0 Rev. 0 Rev. 0 Rev.

1 Rev. 0 Rev. 0 Rev. 0 Rev. 0 Rev. 0 Rev.

1 Rev. 0 Rev.

1 Rev. 0 Rev. 25 Rev. 14 Page 13

References 3.

"Updated Final Safety Analysis Report",

Revision 9,

Palo Verde Nuclear Generating Station Units 1,

2, and 3,

Docket Nos.

STN 50-528/529/530 (including SARCNs 3807 and 98-F033).

PVNGS Procedure 40EP-9EO09, Revision 6, "Functional Recovery".

PVNGS Calculation 13-NC-ZC-206, Revision 3, "Loss of Coolant Accident Containment Pressure and Temperature Analysis".

PVNGS Calculation 13-MC-NC-006, Revision 0, "Nuclear Cooling System Hydraulic Calculation".

PVNGS Calculation 13-MC-WC-303, Revision 1, "WC and NC Relief Valve Sizing".

PVNGS Calculation 13-NC-ZC-232, Revision 5, "Loss of Coolant Accident Pressure and Temperature Containment Analysis for Limiting Case".

PVNGS Study 13-MS-A99, Revision 1, "Evaluation of Thermal Transient Effects on Closed Spaces Between Isolation Valves".

PVNGS Calculation 13-MC-WC-302, Revision 1, "Normal Chilled Water System (WC), MOV Differential Pressure",

page 29b, Section D: Emergency Conditions and Table 1.

10.

12.

PVNGS Calculation 13-JC-HC-202, Revision 1, "Containment Pressure Wide Range (PPS Input) Instrument (HCx-P-352x, x = A, B, C, D) Setpoint and Uncertainty Calculation", Figure 2 and Section 5.1.

APS letter to NRC 102-03855-JML/AKK/JRP, "Response to NRC Generic Letter 96-06", 120 day response, dated January 28, 1997.

APS letter to NRC 102-03943-JML/AKK/JRP, "Supplemental Response to NRC Generic Letter 96-06", dated May 30, 1997.

APS letter to NRC 102-04130-JML/SAB/RMW, "Response to Request for Additional Information Regarding

Response

to Generic Letter (GL) 96-06, "Assurance of Equipment Operability and Containment Integrity During Design Basis Accident Conditions" for the Palo Verde Nuclear Generating Station, Unit Nos. 1, 2, and 3.", dated June 04, 1998.

Page 14

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