ML19331C291
| ML19331C291 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 08/05/1980 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19331C284 | List: |
| References | |
| NUDOCS 8008140509 | |
| Download: ML19331C291 (8) | |
Text
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d, UNITED STATES NUCLEAR REGULATORY COMMISSION n
WASHINGTON, D. C. 20555
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1 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO.19 TO LICENSE NO. NPF-4 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION, UNIT N0. 1 DOCKET NO. 50-338 Introduction This Safety Evaluation Report related to Amendment No.19 to Facility Operating License NPF-4 for the North Anna Power Station, Ur'.1 (NA-1), addresses two license conditions. One of the license conditions required certain actions be completed by the Virginia Electric and Power Cogany (the licensee). The other license condition corrects an administrative error in the paragraph notation of a license condition.
These license conditichs are addressed below in our safety evaluation which provides the basis for removal of License Condition 2.D.(3).g and the redesignation of License Condition 2.D.(3).0, as issued in Amendment No.16 (December 28, 1979) to License Condition 2.D. (3).r.
In addition, this safety evaluation addresses changes to the Technical Specifications as requested by the licensee and/or the NRC. The basis for our approval of these changes to Facility Operating License-NPF-4 and the Technical Specifications is provided below.
Definition for Operability of Safety Systems By letter dated April 10, 1980, we requested the licensee to submit proposed changes to the North Anna Unit 1 Technical Specifications which would clarify the meaning of the term Operable as it applies to the single failure criterion for reactor safety systems.
We stated that a reactor safety system, or any cogonent thereof, shall be defined to be Operable when it is capable of performing its specified,
function (s).
Implicit in this definition is the assumption that all necessary instrumentation, controls, normal and emergency electrical power sources, cooling or seal water, lubrication and other auxiliary equipment are capable of performing their related support function (s) necessary to assure that a reactor safety system or any cogonent thereof will perform its safety related function (s).
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i I By letter dated May 16, 1980, the licensee proposed changes to the NA-1 Technical Specifications in response to our April 10, 1980
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request. The licensee proposed changes for clarifying the inter-action of the Limiting Conditions for Operation of a support system and for operation of equipment.
A We have reviewed the licensee's proposed Technical Specifications for
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conformance to the guidelines of our April 10, 1980 request to the j
i licensee, and we find these changes to be acceptable.
Emergency Core Cooling System (ECCS) Surveillance Reouirements By letter dated July 3,1980, the licensee proposed changes to the l
NA-1 Technical Specifications which would change the ECCS l
Surveillance Requirements. The proposed changes are in response i
to our reconinendations that the NA-1 Specifications be consistent with the format and content of the NA-2 Specifications as well as all other Standard Technical Specifications.
Our request to change Specification 4.5.2.g will require that ECCS i
subsystems be demonstrated operable by verifying that the manual valves listed in Specification 4.5.2.g.2 are locked and tagged in the yli proper position for safety injection following maintenance or repositioning i
of the valves.
In addition, we requested that a Specification 4.5.2.h be added to the I
NA-1 Technical Specifications. The proposed Specification 4.5.2.h requires demonstration of ECCS subsystem operability by requiring i
a flow balance test following the cmpletion of modifications to the ECCS subsystems which alter the s osystem flow characteristics. The
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proposed Specification 4.5.2.h is identical in format and content to l
the currently approved NA-2 Specification.
We find that the above proposed changes increase the margins of safety associated with the ECCS and its subsystem. Therefore, we find the licensee's proposed changes to be acceptable.
Technical Specification Trip Setpoint Values By letters dated September 29, 1978 and October 13, 1978, the licensee f
responded to our letter dated March 15, 1977. Our March,1977 letter requested documentation of explicit technical specification trip setpoint values for the reactor protection system (RPS) and engineered safety features actuation system (ESFAS) at NA-1.
F This requirement for documentation of technical specification trip l
setpoint values was so stipulated in License Condition 2.D.(3)g i
of Amendment No. 3 to Facility License NPF-4 issued on April 1,1979.
The licensee's September 29, 1978 letter provided a report entitled
" Westinghouse Reactor Protection System / Engineered Safety Features Actuation System Setpoint Methodology", and a listing of protective functions and the setpoint allowances (drift and calibration error) used in determining the reactor trip setpoint values for the RPS and ESFAS at NA-1.
THe licensee's October '!
1978 letter proposed changes to the NA-1 Standard Technical apecificaticns for conformance with the methods described in the above mentioned Westinghouse report and Westinghouse Standard Technical Specifications.
Also in a letter dated December 12, 1978, the licensee identified a proposed change to Note 3 of the NA-1 Technical Specification Table 2.2-1.
This change specifies that the maximum trip setpoints
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for the overpower and overtemperature AT reactor trips shall not exceed the computed trip setpoints by more than a two percent span.
The report entitled " Westinghouse Reactor Protection System / Engineered Safety Features Actuation System Setpoint Methodology", details the methods used to determine setpoints and setpoint allowances for the RPS/ESFAS instrumentation at NA-1.
The Westinghouse report describes the assumptions used in determining channel breakdown values for various instrument channel components in 4
the RPS/ESFAS. ' A basic assumption in the Westinghouse technique is that error components may act independently when determining the overall allowance for each setpoint. This assumption allows the use of statistical summation in place of a rigorous additive summation.
For parameter assumptions that are known to be interactive, however, additive summation is used. The benefit of this technique is that it provides an increased margin in the total setpoint allowance.
Using this methodoloy, error components for a channel are combined statistically into groups of components that prove to be statistically independent. Error components that are not independant are added arithmetically in groups. The groups of components are then combined systematically according to the independent effects of each group.
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The methodology described in this Westinghouse report has been documented in other Westinghouse reports (WCAP-9180 and WCAP-8567) that have been l
approved by the NRC staff. Also, the Instrument Society of America l
has recognized the use of these statistical techniques for determining safety-related instrument setpoints.
i The Westinghouse report also describes a new method for verification of setpoints that is also used at NA-l.
This method uses a four-l column table that allows for rack and sensor "as measured" para-meter drifts.
This table reflects both the 31-day and the 18-month surveillance requirements.
"? have reviewed the Westinghouse methodology and we find it to be i reliable statistical technique for determining the total error in t!.4 NA-1 RPS/ESFAS reactor trip setpoints.
We have reviewed the licensee's October 13, 1978 proposed NA-1 Technical Specifications and verified that the RPS/ESFAS trip setpoints (TS Tables 2.2-1 and 3.3-4) are in conformance with the Westinghouse report and the Westinghouse Standard Technical Specifica-tions and are acceptable when corrected as follows.
The correction includes:
(1) An addition of a three percent margin to the steam generator water level -- low-low s,etpoints to be consistent with the NA-2 Technical Specifications as evaluated in Supplement No.10 to the Safetv Evaluation Report for NA-2.
(2) A change to the pressurizer trip from a coincident trig of j
pressure with level to pressurizer pressure -- low-low to conform to IE Bulletin 79-06A. This change has been provided in the NA-1 Technical Specifications in response to the IE Bulletin in Amendment 16 to Facility Operating License NPF-4 issued December 28, 1979.
(3) For steam flow in two steam lines -- high coincident with Tavg -- low-low or steam line pressure -- low (Table 3.3-4) the under allowable value for Tavg should be changed from
>541 to >542 degrees Farenheit.
We have discussed items (1) and (3), as noted above, with the licensee.
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The licensee agrees that these changes should be made to the NA-1 Technical Specifications.
These changes, as noted in items (1) and (3) are hereby incorporr.ced in the NA-1 Technical Specifications.
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l i l The licensee's proposed change to Note 3 of Technical Specificatiori Table 2.2-1 described in its letter of December 12, 1978 has been determined to be necessary. The proposed change is from "4 percent" to "2 percent span" by whicn the maximum trip poirts for the overpower and overtemperature AT reactor trips shall not exceed the coguted trip points. This revision to Note 3 is consistent with the setpoints and allowables presented in the Westinghouse Standard Technical Specifications for North Arna 1 and 2.
We find this proposed change to be acceptable.
Based upon our review, verification and approval of the trip setpoint values for the NA-1 RPS/ESFAS Technical Specification (Tables 2.2-1 and 3.3-4) and the corrected changes as noted above, we find that License Condition 2.D.(3).g is no longer required. Therefore, Facility Operating License No. NPF-4 is hereby amended by ramoving License Condition 2.0.(3).g.
_AuM.tary Feedwater Surveillance Recuirements By letter dated September 28, 1979, we notified the licensee of our then Bulletins and Orders Task Force reconnendations regardirg certain modifications and procedural changes to the NA-1 auxiliary feedwater (AFW) system and its supporting systems. These reconnendations resulted from a reliability study of the NA AFW system. Amon5 these recorrendations was a requirement that, following an extanded cold shutdown, a flow test should be performed to verify the normal flow path from the primary water source to the steam generators.
By letters dated Cecember 28, 1979 ad February 25, 1980, the licensee proposed to add Specification 4.7.1.2c to the surveillance requirements for the auxiliary feedwater system.
Proposed Specification 4.7.1.2c will require that prior to entry into Mode 3 (hot standby) from Mode 4 (hot shutdown) following operation in Mode 5 (cold shutdown) each auxiliary feedwater pump will be started and deliver water to its l
associated steam ger:erator.
In our discussions with the licensee regarding this matter, the licensee's l
proposed change was revised to read:
" Prior to entry into Mode 3 (hot standby) following Mode 5 (cold shutdown) performance of a flou test of each auxiliary I
feedwater pump to verify the normal flow path from the I
condensate storage tank through the pump to its associated steam generator is required."
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The licensee's revised change is in accordance with the Bulletins ard Orders Task Force recomendation for verifying that the AFW flow prch from the primary water source to the steam generator (s) is available prior to power operations after a cold shutdown.
Therefore, we find the licensee's revised change to be acceptable.
The licensee also proposed a specification change to the surveillance requirements for the turbine driven AFW pu.:p.
This change is based on plant operating experience at the NA-1 plant. The original specification required the turbine driven pump to be tested at a main steam pressure greater than 835 pounds per square inch gauge. This was to allow entry into Mode 3 (hot standby) without violating a limiting condition for operation; namely, that the turbine driven AFW pump must be opercble during all phases of Modes 1, 2 & 3.
l However, during Mode 1 (power operation) at higher power levels, 835 i
pounds per square inch gauge was not available and the pump could not be tested in accorda".ce with the surveillance requirements. The licenser. proposed to delete the steam pressure requirement such that the turbine driven AFW pump can be tested during all power levels of Mode 1.
The deletion of the minimum steam pressure necessitates an exemption to the action statement for the operability of the steam driven pump such that heatup into Mode 3 can be accomplished without violating the limiting conditions for operation. The surveillance test will now be performed when heating up in Mode 3 as soon as sufficient i
steam pressure is available to achieve AFW pump flow and discharge pressure specified in the NA-1 Technical Specification surveillance 4
requirements.
Since these changes result in more flexibility for testing the turbine driven AFW pump including testing in Mode 1 without diminishing safety, (when the pumps would be needed most if called upon) we find the proposed change to be acceptable.
Addition of Hydraulic Snubbers (Table 3.7-4)
By letter dated March 31, 1980, the licensee proposed changes to the NA-1 Technical Specifications. The proposed change would revise Table 3.7-4, " Safety Related Hydraulic Snubbers". The revi:. on reflects the addition of hydraulic snubbers resulting from design changes to several safety-related systems.
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. A total of eleven hydraulic snubbers were added to safety-related systems. Specifically, the proposed change will add hydraulic snubbers numbered 800, 801, 802, 803, 804, 450, 233A, 571, 280, 281, and 282 to Table 3.7-4.
This addition was necessary to t
bring the affected safety systems into conformance with the NA-1 FSAR comitments.
However, the addition will not change any systems as described in the FSAR.
Since a properly functioning hydraulic snubber will not offer significant resistance to steady state loadings, its existence will cause no change in steady state system stress patterns. However, the snubber acts as a strut under dynamic loadings and will reduce stress level in general during a dynamic event. The addition of the hydraulic snubbers, as specified above, to Table 3.7-4 will ensure their proper functioning by in-service inspection and testing and enhance plant operating safety. Therefore, we find the proposed changes to be acceptable.
Deletion of Hydraulic Snubber (Table 3.7-4)
By letter dated June 2,1980, the licensee proposed changes to the NA-1 Technical Specifications. The proposed change would remove hydraulic snubber No.103 from Table 3.7-4, " Safety Related Hydraulic Snubbers". The proposed change reflects the licensee's reanalysis and redesign of appropriate supports in the Safety Injection System of the Safeguards Area.
The licensee's reanalysis and redesign indicates that the support location where hydraulic snubber No.103 is located is no longer required. The licensee's reanalysis using existing methods including applicable Amplified Response Spectrum curves yield acceptable results.
Appropriate portions of the safety-related system were reanalyzed using enveloped Amplified Response Curves and the margin of safety will not be reduced by removing hydraulic snubber No.103. Therefore, we find this change to be acceptable.
By letter dated Ju e 2,1980, we provided the licensee confirmation of our emergency authorization for a change in the NA-1 Technical Specifications for deleting snubber No.103 (Table 3.7-4).
Our evaluation, as provided above, formally documents the basis for the authorization provided in our letter to the licensee on June 2,1980.
Administrative Chances In Amendment No.17 (May 19,1980) to Facility Operating License NPF-4 changes were made in the NA-1 Technical Specifications to reflect the licensee's new management organizatica which became effective April 1, 1980.
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l By administrative error, Table 6.2-1, " Minimum Shift Crew Conposition",
as issued in Amendment No.17, stipulated the requirements for NA-1 plant operation only. The appropriate Table 6.2-1, which includes the Minimum Shift Crew Composition for multi-unit plant operation at NA-1 and NA-2 and already approved for the NA-2 Technical Specifications is also necessary for the NA-1 Technical Specifications. The correct a
Table 6.2-1 is hereby incorporated in the NA-1 Technical Specifications.
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In Amendment No. 16 (December 28, 1979) to Facility Operating License NPF-4, the condition for the secondary water chemistry program was incorrectly identified as "new paragraph 2.0.(3).0".
The correct paragraph is +ification for this condition should read "new paragraph 2.D.(3).r".
The correct paragraph identification is hereby added to Facility Operating License NPF-4.
Environmental Consideration We have determined that the amendment does not authorize a change in effluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. Having made this determination, we have further concluded that the amendment involves an action which is insignificant from the standpoint of environmental impact and, pursuant to 10 CFR Sl.5(d)(4), that an environmental impact statement or negative declaration and environmental impact appraisal need d
not be prepared in connection with the issuance of this amendment.
Conclusion We have concluded, based on the considerations discussed above, that:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities will be conducted in compliance with the Comission's regulations and the issuance of this amendment will.not be inimical to the common defense and security or to the health and safety of the public.
f Dated: August 5,1980
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