ML18139C130

Application for Amend to Licenses DPR-32 & DPR-37, Changing Tech Specs to Restore Core Thermal Limits & Overpower Change in Temp Setpoints to Values Consistent W/ 100% of Thermal Design Flow.Safety Evaluation Encl

ML18139C130
Person / Time
Site:	Surry
Issue date:	11/22/1982
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton, Varga S Office of Nuclear Reactor Regulation
Shared Package
ML18139C131	List: ML18139C130 ML18139C132
References
619, NUDOCS 8211290731
Download: ML18139C130 (8)

Text

,_

'VIRGINIA ELECTRIC A.ND POWER COM.PA.:NY RrcE:li.1o:i:-.J:?, Vrno:r:N IA. 23261 W. L. STH,'ART VICE PRES IDEKT November 22 1 1982 NUCLEAR OPERATIONS Mr. Harold R. Denton, Director Serial No. 619 Office of Nuclear Reactor Regulation FR/NAS:bjc I Attn: Mr. Steven A. Varga, Chief Docket Nos. 50-280 Operating Reactors Branch No, 1 50-281 Division of Licensing License Nos. DPR-32 U.S. Nuclear Regulatory Commission DPR-37 Washington, D.C. 20555 Gentlemen:

Af~EKDMEKT TO OPERATING LICENSE DPR-32 AJ{D DPR-37 SURRY POWER STATION UNITS NO. 1 AND NO. 2 PROPOSED TECHNICAL SPECIFICATIO~S CHANGES Pursuant to 10CFRS0.90, the Virginia Electric and Power Company hereby requests an amendment, in the form of changes to the Technical Spec~:.ications, to Operating Licenses DPR-32 and DPR-37 for the Surry Power Station, Unit Nos: 1 and 2. The proposed changes are enclosed.

In August of 1977 (Letter from C. M. Stallings (Vepco) to E. G~

Case (KRC), Serial No. 344, August 9, 1977) Vepco provided the justif~cation for operation of Surry Units 1 and 2 with substantial levels of steam generator tube plugging. The submittal addressed the

i.mpact on non-LOCA accident analyses of steam generator tube plugging levels of up to 40% with consequent Reactor Coolant System (RCS) flow reductions to as low as 90% of the thermal design. flowrate considered in the Final Saf e.ty Analysis Report (FSAR). Also, a revised set of *core -

thermal operating limits a~d corresponding overtemperature and overpower

~T setpoints were submitted, consistent with the assumption of 90% of design flow.

As yo~ kno~, Vepco undertook an extensive steam generator repair program in 1979-1980, resulting in total replacement of the steam generator tube bundle regions for beth Units 1 and 2. As a result, full thermal design flow has been re-established as the appropr:iate de. 9,ign

• basis for thos~ units. However, since the completion of the re~~ir program, Vepco has retained the conservative core limits and setpoints submitted in.1977.

~e hereby request a revision to the Technical ~pecifications to restore the core thennal limits and overternperature and overpower bT setpoints to values consistent with 100% of thermal design flow. The proposed limits are identical to those submitted by letter from C. M. * -*

Stallings (Vepco) to KR. Goller (NRC), Serial No. 458, March 12, r97s:*

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2. +11 ~v&}r,q3J

r e VIRGINIA ELLCTEIC AND Po .....'1cH CoHPA1'-Y TO, Mr. Harold R. Denton 2 in conjunction with our Reload Safety Evaluation for Surry tJnit 2, Cycle

2. These were the previously applicable limits, in effect prior to our August 1977 submittal; consequently, no additional accident analyses are required to support this change. Furthermore, in addition to the normal conservatisms standard to Westinghouse DNB analyses, these limits reflect a fuel densification power spike penalty which was subse,quently eliminated.

• It should also be noted that these limjts include an allowance for an increase in the enthalpy rise hot_ channel factor ba!5_ed on the*

expression:

F:H = 1.55 (1 + 0.2 (1 - P))

where Pis fraction of rated power. At a later time, we will be N submitting the technical justification to support an increase in FbH to the value:

N Ftill.= 1.55 (1 + 0.3 (1 .,. P))

At that time, revised core thermal limits and, if necessary, eore protection setpoints will also be submitted for approval, This request has been reviewed and approved by the Station Nuclear Safety and Operating Committee and Safety Evaluation and Control Staff.

L: *has been determined that this request does not involve any unreviewed safety questions as defined in 10CFR50.59.

We have eval~ated this request in accordance with the criteria in 10CFR170.22, It has been determined that this request requires a Class III Amendment fee. Accordingly, a voucher cl1eck in the amount of

$4400.00 is enclosed in payment of the required fee, In order _to restore plant operational flexibility in a timely manner, we request approval of this change by January 1983, 'in order to allow for implementation prior to or during ~tartup of Surry 1, Cycle 7 and Surry 2, Cycle 7.

Jit(t:J~

\ -

W. L. St-ewart .

J Attachments (1) Safety Evaluation (2) Proposed Technical Specifications Changes (3) Voucher Check for $4400.00.

cc: Hr. James P. O'Reilly, Regional Administrator Region II

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• RICHMOND, VIAOINIA L

DANK OF VIRGINIA .

PAY

'DK l,nECK NO. " :p/\ii!

• 1 YHtDOn No. I AMOUNT

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25i ,,',\ '11)3k\J1,.

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.o COMMONWEALTH OF VIRGINIA)

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CITY OF RICHHO:t-.'D )

The foregoing document was acknowledged before me, in and for the City and Col!l!llonwealth aforesaidp today by W. L. Stewart, ~ho is Vice President-Nuclear Operations, of the Virginia Electric and Power Company_._ He is duly authorized to execute and file the foregoing document i.n behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.

Acknowledged before me this-?<<~' day of ?J~~ 0 19 ~.;a._

My CoI!l!llission expires: 19 ~~

Notary Public

~-* (SEAL}

H2/004

e Attachrnent 1

~afety Evaluation Full Flo~ Thermal Limits - Surry Units 1 and 2 During the period from 1975 to 1979 Virginia Electric and Power Company faced substantial*difficulties vith steam generator tube degradation and leaking at Surry Power Station, as did a number of other***utilities throughout the world. *As a result cf this problem, a substantial percentage of the steam generator tubes on both Units 1 and 2 ~ere plugged over the course of several refueling and maintenance outages.

This tube plugging had several impacts. The loss of steam generator

)

tube heat transfer area resulted in some degradation in secondary steam pressure and loss of overall plant thermal efficiencyo Also, the.plugging changed the hydrc1ulic resistance (pressure drop characteristics) of the primary side of the steam generators. This increased pressure drop had two effects on the plant safety analyses. First, the transient hydrodynamics of the large break Loss Of Coolant Ac~ident (LOCA) were affected; as a result, a series of-reanalyses of the LOCA at various levels of plugging were required. Secondi the amount of margin between actual Re2ctor Coolant System (RCS~ flow rates arid the thermal design flo ...1 rc1tes assumed in the safetv analysis was continually

~

r~ciuced as tube plugging levels increased *.

In Au~ust of 1977, (Reference 1) Vepco provided the justification for operation of the Surry units with steam generator tube plugging l~vels_ of up to

--* 40%, along "With associated reductions in RCS flow rate to as low as 90% of the thermal desigr. flow rate considered in the Final Safety Analysi5 Report (FSAR).

This suh~ittal addressed the i~pact of this tube plug~jng level on non-LOCA

&ccidents. (The impact of plugging levels of up to 28% on the large-:break_ LOCA.

. ;;;as addressed in B e

• seperate. submittal - see Reference 2.) The Ref erenci:: l .

submittal also provided a revised set of core thermal operating limits and corresponding overtemperature and overpower ~T setpoints consistent with the assumption of 90% of design flow. The safety analyses presentea in the submittal 'l:iere based on these "lo,.., flown setpoints.

From 1979 to 1981>> Vepco undertook an extensive steam generator rep2ir program? resulting in total replacement of the steam generator tube bundle regions for both Units 1 and 2. Startup measurements for subsequent cycles have confirmed that system flow rates for both units.are vell in excess of the thermal design value (measured values are from 10 to 15% in excess of thermal design flo~ (see References 3 to 5)). Steam generator performance following the replacement programs has been good; no tube plugging has been requi,recJ to date. Consequently, the 100% thermal design flow rate is again an appropriate basis for determining core thermal operating limits*and protection setpoints.*

Hov~yer, since the completion of the repair program, Vepco has retained the conservative core limits submitted in Reference 1.

The revised limits and setpoints provided in*the proposed Technical Sped fications changes are consistent with the FSAR assumption of 100%. c:if thermal design flow, and are identical to those submitted by Reference 6.

These were the previously applicable limits in effect prior to the Reference 1 submi tt81. The basis for these limits is discussed ir. Reference '7; as such, the limits reflect an explicit, conservative occounting for the effect of a derrsification-induced power spike on DNBR. This conservatism is in addition to the normal conservatisms standard to Westinghouse DNB Analyse~ *.. -;,"

Since the proposed limits and setpoints sre consistent with those previously analyzed, no accident reanalysis is required for the proposed change. As has been previously identified, operation with the revised limits 2

( e

• and setpoints does not pose an unreviewed safety question as defined in 10 CFR 50.59 or invalidate any existing safety analyses for Surry Units 1 and 2.

/

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. e References 1 *. Letter from C. M.. Stallings (Vepco) to E.G. Csse (NRC), .Serial No. 344,

/ August 9, 1977.

2. Letter from C. M. Stallings (Vepco) to H. R. Denton (NRC), Serial No. 736, December 26, 1978.

3. Letter from W. N. Thomas (Vepco) to H. R. Denton (NRC) Serial No. 873, October 29, 1980, "Surry Power Station, Unit 2, Cycle 5 Startup Physics Test Repcrt. 11

4. Letter from R.H. Leasburg (Vepco) to H. R. Denton (NRC) .Serial No. 541, September 9, 1981, "Surry Power Station Unit l, Cycle 6 Startup Physics Test-Report."

5. Letter from R.H. Leasburg (Vepco) to H. R. Denton (NRC), Serial No. 073, March 8, 1982, "Surry Power Station, Unit 2, Cycle 6 Startup Physics Test Report. 11 6~; Letter from C. M. Stallings (Vepco) to K. R. Goller (NRC),Serial No. 458,,

_,../ ,~ci,rch 12, 1975.

7. Fuel Densification - Surry Power Station Unit 1, 11 WCAP-8012 (Proprietary),

WCAP-8013 (Non-proprietary), December 1972.

132eNAS62

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