ML18153B725

Provides Addl Info Re 880411 Tech Spec Change Request Concerning Inservice Insp & Testing Requirements.Info Covers Visual Insp of Safety Injection Sys Piping in Valve Pit & Evaluation of Acceptance Criteria for Main Steam Valve

ML18153B725
Person / Time
Site:	Surry
Issue date:	05/12/1989
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
89-304, NUDOCS 8905240062
Download: ML18153B725 (10)

Text

e e VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 23261 May 12, 1989 United States Nuclear Regulatory Commission Serial No.89-304 Attention: Document Control Desk NO/ETS Washington, D. C. 20555 Docket Nos. 50-280 50-281 License Nos. DPR-32 DPR-37 Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 TECHNICAL SPECIFICATION CHANGE REQUEST INSERVICE INSPECTION AND TESTING SUPPLEMENTAL INFORMATION By letter dated April 11, 1988, we submitted proposed changes to the Technical Specifications to meet the new inservice inspection and testing requirements for nuclear components. In an April 11, 1989, telephone conference call with your staff additional information was requested to complete your review of our lnservice Inspection Technical Specification request of April 11, 1988. This letter provides the requested information. Attachment 1 is the revised Technical Specification pages.

Page 4.2-4 reinstates, per your request, the requirement to visually inspect the Safety Injection System piping in the valve pit every refueling outage and Page 4.2-5 items 1.3 and 1.4 have been renumbered to accommodate the above requirement.

Attachment 2 is a summary of the engineering evaluation that supports the +3% "as-found" acceptance criterion for Main Steam Safety Valve testing.

This supplemental change has been reviewed and approved by the Station Nuclear Safety and Operating Committee and the Safety Evaluation Staff. The change does not impact the previous safety evaluation nor the significant hazards consideration determinations.

• If you have further questions please call.

Very truly yours,

,~)L~\)-

w.L:m~~

Senior Vice President - Power Attachments I

( 8905240062 890512 PDR ADOCK 05000280 p PDC 1

I e e cc: U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. W. E. Holland NRC Senior Resident Inspector Surry Power Station Commissioner Department of Health Room 400 109 Governor Street Richmond, Virginia 23219

e e ATTACHMENT I SUPPLEMENTAL TECHNICAL SPECIFICATION REQUEST INSERVICE INSPECTION AND TESTING REQUIREMENTS

TABLE 4.2-1 SECTION A. MISCELI.ANEOUS INSPECTIONS Required Required Item Examination Examination Tentative Inspection No. Area Methods During 10-Year Intei:val Remarks 1.1 Materials Tensile and capsules shall be removed capsule #1 = First refueling Irradiation Charpy V notch and examined after 10 capsule #2 = At five years SUrveillance (wedge open years. (See Notes 1 and 2) capsule #3 = At 10 years loading) and capsule #4 = At 20 years doslllletry as capsule #5-8 = Are spares necessary to for complemen-insure sur- tary or dupli-veillance cate testing.

1.2 IDw Head SIS Visual Non-applicable This pipe shall be visually piping located inspected at each refueling in valve pit shutdown.

Note 1: 1 year corresponds to 1 year effective full power operation.

Note 2: The results obtained from these examinations shall be used to upjate Figure 3.1-1 as required.

e L_

i I

TABIB 4.2-1 SECTION A. MISCELIANEOUS INSPECTIONS Required Required Item Examination Examination Tentative Inspection No. Area Methods During 10-Year Interval Remarks 1.3 Primal:y Pump See remarks See remarks Examination to be conducted Flywheel in accordance with regulatory position C.4.b of regulatory guide 1.14 Rev. 1, August 1975 1.4 low Pressure Visual and 100% of blades every 5 None Turbine Rotor Magnetic years Particle or Dye Penetrant SECTION B. SENSITIZED STAINLE'SS STEEL 2.1.1 Circmnferential Visual and By the end of the inter- A minimum of 5% of the welds and longitudinal Volumetric val, a cumulative 75% of will be examined every 1-2/3 pipe welds and the circumferential welds years (generally each normal branch pipe in the piping system refueling outage). See connections would have been examined, Transcript of Hearing (pp.

larger than 4 including one foot on any 303--34) and Initial De-inches in longitudinal weld on either cision (p.7, p.10) diameter side of the butt welds 2.1.2 Circmnferential Visual By the end of the inter- A minimum of 50% of the welds and longitudinal val a cumulative 100% of will be examined every 1-2/3 pipe welds and the welds and pipe branch years (generally, each normal branch pipe connections would be refueling outage). See connections examined a minimum of Transcript of Hearing (pp. ci5 three times. 303-304) and Initial De-

~

cision (p.7, p.10) I

e e ATIACHMENT2

SUMMARY

OF ENGINEERING EVALUATION

+/- 3% ACCEPTANCE CRITERION

e e SAFETY EVALUATION

SUMMARY

An evaluation of the impact of an as-found Main Steam Safety Valve (MSSV) setpoint tolerance of +/- 3% on the UFSAR accident analysis has been performed for Surry Power Station. The results show that there is no adverse impact on the accident analysis results in that appropriate de-sign limits continue to be met and no unreviewed safety question as de-fined in 10 CFR 50.59 is created. The results of this evaluation are discussed in more detail below.

The impact of the MSSV tolerance for the design heat load rejection transient (loss of load accident) was examined first. A RETRAN sensi-tivity study was performe_d to examine the effects of mode 1ing the 3% as-found tolerance with respect to the current licensing analyses. The case examined was a 100% load rejection from hot full power with no credit for condenser dumps or atmospheric relief valves. BOC (minimum temperature and maximum power feedback) conditions were assumed. This will be the limiting overpressure event from the standpoint of impact of MSSV setpoints.

The results showed that the proposed main steam safety valve tolerance had no impact on the peak reactor coolant system pressure. The secondary side pressure increased only slightly (less than 0.5%) and was still well within 110% of the system design pressure.

2

. ,' e While the locked rotor event can give higher pressures on the primary side, the overpressure condition is driven by a rapid expansion of RCS water in the core combined with a loss of heat transfer to the secondary in the affected loop. MSSV setpoints would thus not be expected to sig-nificantly impact the primary side results for a locked rotor event. The effect on the secondary side pressure was examined by rerunning the cur-rent licensing analysis with the relaxed MSSV tolerances. The results showed that the secondary steam pressures remain within the limit of 110%

of design pressure.

An evaluation was also performed to assess whether the proposed setpoint tolerance could affect the capability to deliver the auxiliary feedwater

--- --(AFW)'

fl ow required for core decay heat removal fo 11 owing a 1oss of norma 1 feedwater (LONF). The results showed that in the event of a LONF, the decay heat removal requirements would not drive secondary pressures high enough to reduce AFW flow below the design requirement assumed in the safety analysis. Further, the potential increase in post-trip RCS tern--*

perature following a LONF is only a few degrees F and will have no impact on the conclusions of the UFSAR analysis.

A review of the effects of the MSSV setpoint tolerance on the effective-ness of the overtemperature/overpower delta-T protection setpoints was performed. It was concluded that the +/-3% tolerance will have no impact on the ability of the current protection setpoints to protect against DNB, vessel exit boiling or high fuel centerline temperatures.

3

..,1'- l' e

The effects of the 3% setpoint tolerance on the low side (as-found 3%

below design) on the safety analysis are bounded by the assumption that the condenser and atmospheric dump va 1ves function normally. If the lowest setpoint safety valve (1100 psia) were to open 3% low, or 1067 psia, this is 17 psi above the nominal setpoint for the atmospheric steam dumps. The capacity of an atmospheric dump valve is approximately the same (about 3% higher) as the capacity of the lowest setpoint safety valve. The relaxed tolerances will not significantly increase the po-tential for post-trip actuation of the MSSV 1 s. This potential will remain low due to the normal availability of the condenser steam dump system, the availablity of the atmospheric PORVs, the unlikelihood of the MSSV lift point being at the bottom of its tolerance band, and the margin be-tween the actuation pressures for these components.

In conclusion, the current practice of using a+/- 3% as-found tolerance on the Main Steam Safety Valve (MSSV) as-found pressure relief setpoint does not create an unreviewed safety question as defined in 10 CFR 50.59.

Specifically:

1) Use of the 3% tolerance does not increase the probability of occurrence of any accident previously evaluated in the UFSAR, since no new accident initiators result from the MSSV pressure relief setpoints. Also, the consequences of the UFSAR accidents are not increased since primary and secondary 4

pressures remain within appropriate design limits (i.e., less than the 110% overpressure limit) during the postulated accidents.

2) The proposed tolerances will not create the possibility of an accident or malfunction of a different type than previously evaluated. No design change to the valves is involved. The proposed Technical Specification change merely establishes a realistic and workable tolerance on the as-found relief setpoints for these valves. An evaluation has shown that relief setpoints as much as 3% above the nominal setpoint will not compromise the ability of the auxiliary feedwater system to deliver flow to the secondary side under normal or accident conditions.

3) The margin of safety as defined in the basis for any technical specification is not reduced, since the results of the accident analyses remain within their appropriate acceptance criteria.

5