ML20237J690

Summary of 870730 Meeting W/Util & C-E in Phoenix,Az Re Util 870629 Submittal of Cycle 2 Reload Analysis & Proposed Tech Spec Changes.List of Attendees,Licensee Draft Responses & Addl Questions Raised by NRC Encl

ML20237J690
Person / Time
Site:	Palo Verde
Issue date:	08/19/1987
From:	Licitra E Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8708260265
Download: ML20237J690 (10)

Text

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AUG 131987 Docket Nos.: 50-528 and 50-529 l

LICENSEE: Arizona Public Service Company l

l FACILITY: PALO VERDE, UNITS 1 AND 2 i

SUMMARY

OF MEETING REGARDING CYCLE 2 RELOAD SUBMITTAL A meeting was held on July 30, 1987 in the licensee's Phoenix office with representatives of the licensee to discuss Palo Verde Unit 1, Cycle 2 r reload. The purpose of the meeting was to discuss the licensee's June 29, l

1987 submittal relating to the Cycle 2 reload analysis and the proposed technical specification (TS) changes for Cycle 2. Enclosure 1 provides the list of attendees and the meeting is summarized as follows:

i l S"MMARY The reload analysis and the proposed TS changes consisting of 15 packages, were submitted by two letters dated June 29, 1987. At the meeting, the licensee gave a summary discussion for each of the 15 packages; the first 12 are directly related to the reload while the last 3 (concerning the Axial Shape Index, Refueling Actuation Signal trip value, and administrative changes to the bases sections for three specifications) are not needed before Cycle 2 l startup.

The licensee stated that a revised figure will be submitted for the proposed Shutdown Margin TS change to be consistent with its proposed temperature i dependent Shutdown Margin currently under review by the staff. For the '

proposed change to the TS on Axial Shape Index, the staff stated that i

additional information will be required since the package provided didn't l identify how changes to the Axial Shape Index would be controlled or how they would be reviewed by the NRC.

By letter dated July 22, 1987, the staff had requested additional information on the reload submittal., At the meeting, the licensee presented draft responses which are provided as Enclosure 2. The staff stated that the proposed responses would provide the requested information, except for the draft response to Question 3 regarding an inadvertent boron dilution event.

The licensee will formally submit the responses and provide additional information for its response to Question 3. The licensee will also provide additional information for the reload analyses to clarify the use of the variable overpower trip in the steam line break analysis.

Three additional questions were raised by the staff at the meeting, which are provided as Enclosure 3. The licensee stated that it would provide a response to these questions along with the other responses.

8708260265 070819 PDR ADOCK 05000528 P PDR

CD -0 6 At the end of the meeting, the licensee provided infonnation relating to Palo.

Verde Unit 2 reload. The end of Cycle 1 for Unit 2 is scheduled for  ;

February 21, 1988 with restart targeted for May 1, 1988. The Unit 2 reload submittal will be made by December 18, 1987 and, for the most part, will be the same us.the reload package submitted for Unit 1.

u The staff concluded that the meeting was helpful in its review of the licensee's reload submittal.

Original signed by: =I E. A. Licitra .

E. A. Licitra, Project Manager Project Directorate V  ;

Division of Reactor Projects - III, '

IV, V and Special Projects

Enclosures:

1. Meeting Attendees

2. Draft Responses to NRC Questions

3. Additional NRC Questions cc w/ enclosures:

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OFFICIALRECONDCOPY t

Mr. E. E. Van Brunt, Jr.

Arizona Nuclear Power Project Palo Verde CC: l Arthur C. Gehr, Esq. Kenneth Berlin, Esq.

Snell & Wilmer Winston & Strawn 3100 Valley Center Suite 500 Phoenix, Arizona 85073 2550 M Street, NW Washington, DC 20037 Mr. James M. Flenner, Chief Counsel Arizona Corporation Commission Ms. Lynne Bernabei 1200 West Washington Government Accountability Project Phoenix, Arizona 85007 of the Institute for Policy Studies 1901 Que Street, NW Charles R. Kocher, Esq. Assistant Washington, DC 20009 Council James A. Boeletto, Esq.

Southern California Edison Company .Mr. Ron Rayner P. O. Box 800 P. O. Box 1509 Rosemead, California 91770 Goodyear, AZ 85338 Mr. Mark Ginsberg Mr. Charles B. Brinkman, Manager Energy Director Washington Nuclear Operations Office of Economic Planning Combustion Engineering, Ir. . j and Development 7910 Woodmont Avenue Suite 1310 l 1700 West Washington - 5th Floor Bethesda, Maryland 20814 {

Phoenix, Arizona 85007 Mr. Wayne Shirley Assistant Attorney General Bataan Memorial Building Santa Fe, New Mexico 87503 Mr. Roy Zimmerman i U.S. Nuclear Regulatory Commission P. O. Box 239 Arlington, Arizona 85322 Ms. Patricia Lee Hourihan 6413 S. 26th Street Phoenix, Arizona 85040 ,

i Regional Administrator, Region V U. S. Nuclear Regulatory Commission 1450 Maria Lane Suite 210 ]

Walnut Creek, California 94596  !

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ENCLOSURE 1 PALO VERDE-UNIT 1, CYCLE 2 RELOAD MEETING ATTENDEES-July 30, 1987 Name Affiliation Manny Licitra NRC/NRR/PDS Larry Kopp NRC/NRR/RSB Gerald Sowers APS/ Nuclear Fuel Paul Crowley APS/ Nuclear Fuel Richard Bernier APS/ Nuclear Fuel Mike Nelson APS/ Nuclear Fuel =

Majid Saba APS/ Nuclear Fuel

.Ron Land APS/ Nuclear Fuel APS/ Nuclear Licensing Lonnie Marker APS/ Nuclear Safety Group Bruce Miller Nancy Turley APS/ Nuclear Fuel CE J. F. Church 1

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ENCLOSURE 2 NRC QUESTIONS ON PALO VERDE BRAF;T i

UNIT 1, CYCLE 2 RELOAD AND ASSOCIATED TECH SPEC CHANGES I

1.a. Q. The justification for the proposed change to delete the Tech Spec allowance for. degraded resistance temperature )

detector (RTD) response times beyond 8 seconds' mentions l that if a RTD response time is greater than 8 seconds, the associated CPC channel must be declared inoperable until repairs are completed. Where is this requirement stated in the Tech Spee?

A. Technical Specification 3/4.3, Table.3.3-2, Note ##

states, ... The measured response time of the slowest RTD shall be less than or equal to 8 seconds." If the RTD response time is greater.than 8 seconds, the above

. condition is obviously not met. Failure to meet the response time test implies the equipment is not functioning properly and therefore must be declared inoperable.

b. Q. Since the RTD response times need only be measured once l per 18 months, what will be done about RTD's for which I previous measurements indicate possible degradation  !'

beyond the 8 second response time before the next measurement?

A. An explicit allowance is not provided for RTD response time drift. If the response time measurements indicate a drift is occurring which might exceed the 8 seconds, appropriate action would be considered at that time.

The RTD response times measured to date at Palo-Verde do not indicate a trend that the 8 second response will be exceeded.

2. Q. Explain why the revision to Action Statement 6.b.1 in Technical Specification Table 3 3-1 does not require the LHR margin to be maintained also.

A. The current requirement in Action Statement 6 b.1 in Technical Specification Table 3.3-1 to maintain Technical Specification 3.2.1 is redundant. Either of the two core power distribution monitoring systems, the Core Operating Limit Supervisory System (COLSS) and the.

Local Power Density Channels in the Core Protection Calculators (CPC's) with either CEAC's operable or .

inoperable, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits. The linear heat rate LCO is based on LOCA analyses which j are unaffected by CPC/CEAC operability. Therefore, no change to the LHR Technical Specification is needed for Action Statement 6 b.1.

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3. Q. In view of the increase in critical boron concentration and boron worth for cycle 2, why was the inadvertent boron dilutdon event not reanalyzed?

A. The revised temperature dependent shutdown margin analysis in the proposed Technical Specification amendment bounds both Cycle 1 and Cycle 2 of Unit 1.

4.a. Q. Since thg allowable MTC limit has been increased to

+0.5x10 , why weren't those transients and accidents which are adversely affected by a positive MTC (e.g.

CEA withdrawal, CEA ejection, LOCA) reanalyzed for Cycle 27 A. The Reload Analysis Report (RAR) glyes MTC values for Cycle 2 in Figure 7.0-2 of +0.5x10 delta-rho /deg. F only at zero core power at BOC. At 100% core power, the MTC has a value of 0.0 which is the same value that the Cycle 1 MTC had at 100% power and BOC (see 1). The Cycle 1 MTC value at zero power was Figure

+0.22x107.g delta-rho /deg. F at BOC.

The three transients mentioned in the request (CEA-withdrawal, CEA ejection and LOCA) will thus be unaffected by the MTC when initiated at 100% care power conditions. The CEA withdrawal transient at low power for Cycle 2 is bounded by the CESSAR agalysis. The CESSAR analysis used a MTC of +0.5x10 delta-rho /deg. F.

Note that the RAR only lists events which are not l bounded by the reference cycle or events for which analysis methodology differed from the reference cycle.

The CEA ejection and withdrawal events are most limiting at full power including consideration of a positive MTC at low powers.

4.b. Q. How does the use of an MTC as a function of power level affect the results of these transients and accidents as compared to an MTC as a function of temperature as was used in Cycle 1.

A. There is no affect on the results of the analysis as the MTC is only being more clearly represented as a function of power level instead of temperature.

c. Q. As stated in the letter from A. Thadani (NRC) to J. K.

Gasper (CE Owners Group) dated June 12, 1987, the staff I is concerned about trends in current reload designs

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that may lead to initial MTC values that are less conservative than those that were assumed in the ATWS analyses. Based on the more positive MTC anticipated ,

for Cycle 2, please comment on the justification for  !

the continued applicability of your ATWS analyses.  !

A. In the above referenced letter, the NRC requested additional information from the CE Owners Group concerning the impact of more positive MTC's on the previously performed ATWS analyses. These analyses were performed on a generic basis in the 1970's by CE for the Owners Group. At the time that the ATWS issue  ;

was originally addressed, both the NRC and the [

individual plant owners considered the Owners Groups an l acceptable vehicle by which to respond'to the staff's  ;

I concerns.

ANPP proposes to address the staff's concerns on the  !

validity of the original ATWS analyses generically via i the CE Owners Group instead of separately through individual plant reload analyses. This will allow the Palo Verde Unit 1, Cycle 2 reload to proceed in an orderly and timely manner. Also, utilizing this proposed approach will provide sufficient time to fully  ;

consider the implications of the impact of more )

positive MTC's on the ATWS analyses.

J 5.a. Q. How was acceptable LOCA ECCS performance for both small and large breaks confirmed for up to 400 plugged steam generator tubes?

A. A large number of plugged steam generator (SG) tubes could potentially affect the ability of the Emergency Core Cooling System to limit the consequences of a Loss of Coolant. Accident (LOCA). Tube plugging increases the RCS flow resistance, and decreases the SG. heat transfer area. The increased flow resistance primarily affects the large break LOCA causing a reduction in the core reflood rate which increases the peak clad temperature for the event. The reduction in the SG heat transfer area.primarily affects the small break LOCA ultimately causing greater core uncovery which increases the peak clad temperature.

The large break LOCA is the limiting case for demonstrating conformance to the 10CFR50.46 criteria.

As indicated above, the major effect of steam generator tube plugging on large break LOCA is the increased resistance to flow. The current LOCA analyses assume a conservatively large pressure drop through the steam generators. The assumed value is more than adequate to

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offset the small increase due to tube plugging. The best estimate SG pressure drop with 400 plugged tubes /SG is 39.5 psi vs. 42.0 psi assumed in LOCA analyses. Therefore, the current large break LOCA analyses cover up.to 400 plugged tubes /SG.

Estimates of the increase in peak clad temperature (PCT) are much less than 100 F for a small break LOCA Therefore, the assuming 400 plugged tubes /SG. 4 estimated PCT for PVNGS is less than 1730 degrees which l demonstrates that the small break LOCA PCT is well l within the 10CFR50.46 limit and that the large. break j LOCA PCT remains limiting.

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b. Q. Was asymmetric plugging modeled?.

A. Asymmetric steam generator tube plugging was considered 1 and it was concluded that the symmetric tube plugging ]

analysis implicitly covers unequal plugging between 1 generators, provided no more than 400 tubes are plugged in either steam generator. The basis for this ,

conclusion is as follows: I Tube plugging affects primarily the refill /reflood portion of the large break LOCA transient. Steam generator tube plugging will have no significant effect on the blowdown portion of the transient. The main impact of steam generator tube plugging is to increase the resistance te flow passing through the primary side of the steam generator, thereby inhibiting steam venting from the core outlet plenum to the break. This reduces the refill /reflood rates and increases the peak cladding temperature. With regard to this effect, plugging fewer than 400 tubes in either or both steam generators will reduce the flow resistance and will improve the refill /reflood rates. Consequently, a reduction in the number of plugged tubes in either i steam generator will reduce the peak cladding temperature relative to the analysis results;with 400 plugged tubes in each generator.

c. Q. Since the loop resistance is affected by tube plugging, justify why new hydraulics calculations were not performed for the Cycle 2 LOCA analysis.

A. As discussed in (a) and (b), it was not necessary to-perform new hydraulic calculations for the Cycle 2 LOCA

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analyses. The assumed pressure drop through the steam generators used in the analyses has enough conservatism to provide for 400 plugged tubes per steam generator.

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6. Q. If 400 plugged steam generator tubes are accommodated in Cycle 2, how do the CPC algorithms account-for tube flow?

plugging in the determination of core A. The core flow used by_CPC in its"DNBR calculation is calibrated every shift to a' pump differential pressure or calorimetric flow measurement per Technical note 7. This calibration Specification Table 4.3-1,in accounts for any change flow resulting from plugged steam generator tubes.

7. Q. Please present the calculated k i and 95/95 probability / confidence uncertain (,es for the Palo Verde L

new and spent fuel racks containing 4.05. weight percent U-235 fuel enrichment.

A. An enrichment of 4.30 weight percent U-235 was assumed in the original criticality analyses of the Palo Verde A description of the new and spent fuel racks.

analyses and assumptions utilized as well as the results of the analyses are presented in Section 9.1 of the Palo Verde FSAR. The largest k calculatedovertherangeofconditi$ns.thatwas which were analyzed was 0.777 for the new fuel racks and 0.889 for the spent fuel racks. The associated 95/95 probability / confidence uncertainties for these calculations were not explicitly determined for Palo Verde due to the large margins that exist between the calculated k e 's and the Regulatory Guide requirements.ff It should be noted that using the same methodology at other CE plants has-always yielded an uncertainty of less than 2.6%. Also, an additional conservatism exists for the k 's reported above in l

Unit 1, Cycle 2 since the maxikdm enrichment will be 4.05 weight percent U-235.

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ENCLOSURE 3 3

- I ADDITIONAL NRC' QUESTIONS l

1. A value of 8.5% was used for the' radial peaking. factor in. crease in the CEA drop analysis. How was this obtained and how is it verified to be- .

the largest change obtainable f6r a CEA drop into'a radial configuration l allowed b'y:the Tecls Spec PDIL' transient insertion limit? l

2. The. proposed change to Tech Spec 3.1.2.3 would permit more than one )

charging pump to be operable during Mode 5. please verify that the assumptions,used in the postulated mass addition event' analyzed for supporting ths Low Temperature Overpressure Protection (LTOP) system design remain' valid with this proposed change.

Does the Bases for the Boration Systems have to be modified for this ,

proposed change?

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3. What additional changes have to be made to the Cycle 2 reload analyses l because of the revised Unit 1,2, and 3 Cycle 1 shutdown margin cwve?

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