ML20244A579
| ML20244A579 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/28/1979 |
| From: | Verelli D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20244A577 | List: |
| References | |
| TAC-07811, TAC-08019, TAC-7811, TAC-7815, TAC-8019, NUDOCS 7904120317 | |
| Download: ML20244A579 (8) | |
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UNITED STATES yS NUCLEAR REGULATORY COMMISSION i
..! fij,)-j-WASHINGTON, D. C. 20555
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' MARCH 2 8 1979 Docket No. 50-321 l
LICENSEE:
Georgia Power Company (GPC)
FACILITY:
Edwin I. Hatch Nuclear Plant MEETING
SUMMARY
ON INSERVICE INSPECTION AND TESTING PROGRAM AT HATCH NUCLEAR PLANT UNIT N0. 1 On February 21, 1979 members of the staff and representatives of Georgia Power Company (the licensee) met at the licensee's Hatch site to discuss the Inservice Inspection and Testing Program for Hatch Unit No.1.
The purpose of the meeting was to identify the additional information or further justification needed by the staff to support the licensee's request for relief from ASME Code requirements which he deems impractical.
The licensee's request for relief was submitted by letter dated August 3,1978. A list of attendees is attached as Enclosure 1.
A summary of the results of the meeting and the staff requirements for licensee action and open items for staff action follows:
1.
Inservice Inspection The staff indicated that numerous places of the it ensee's submittal requested relief based on the existence of an inspection program in the facility Technical Specifications.
Examples include relief requests 2.1.2 and 3.1 as they relate to verification of snubber settings.
In such cases a determination must be made as to whether the requirements of the Code are met or exceeded by the Technical Specifications.
The licensee indicated that the alternate inspection exceeds the Code requirements.
The staff recommended that the requested relief be revised accordingly.
The staff indicated that certain of the licensee's requests for re. lief
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not identify particular systems or portions of the system for which lef was requested.
The staff agreed to identify those portions of ue program to which this comment applies; see Enclosure 2.
The licensee agreed that a revised submittal will address this issue.
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Meeting Summary.
The staff indicated that portions of the inspection program requested I
relief based on a generally stated determination of inaccessibility or l
impracticality due to radiation exposure considerations. The licensee's l
revised submittal should provide a definitive technical basis (to l
include quantification of radiation burden) to support any relief request.
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2.
Inservice Testing The staff related the NRC position relative to ASME Code requirements IWV-3520(a), (b), (b)(1) and (b)(2),1974 Edition for check valves.
The staff expects that check valve testing will consist of a full stroke, or, if only limited operation is possible during power operation, the test shall be a partial stroke.
Since the disk position of the valve is not always observable, the staff considers a fluid flow test to be an acceptable alternative.
A valve is considered to be full stroked when the measured flow is at least that which is identified in the plant's safety analyses for the appropriate valve in question. Any less measured flow will be considered as a partial stroke test unless it can be shown that for any less flow the valve disk would be full open against its stop.
The licensee should review his submittal using this guidance and revise it as necessary.
Attached to NRC letter dated November 22, 1976 was the staff's,
" Guidelines for Exdluding Excercising (cycling) Test of Certain Valves During Plant Operation". The staff noted that these guidelines pointed out that if valves, which when cycled, could subject a system to pressure in excess of their design pressures, they should not be tested during power operation.
It is assumed for the purpose of a cycling test that one or more of the upstream (or downstream) check valves has failed unless positive methods are available for determining the pressure or lack thereof on the high pressure side of the valve to be.
cycl ed. The licensee should carefully review his Technical Specifications including the request for Amendment dated October 3,1978 to insure that these guidelines are implemented, where possible.
Should any conflict be identified, the licensee should amend his request for a Technical Specification change.
The staff noted that the NRC differentiates between cold' shutdown and
"'ueling for purpose of Code requirements for valve and pump testing.
V cold shutdown, the staff's intent is that valve testing commence as soon as possible into the cold shutdown mode but no later than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after the shutdown.
The intent of the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period is to provide flexibility of scheduling for those shutdowns which occur during the.
weekend.
Valve testing should continue during the shutdown _ until complete or until plant startup and return to power.
Any testing not completed at one cold shutdown should be performed during subsequent cold shutdowns before the next refueling.
All valves identified to be tested at cold.
shutdown are expected to be tested at refueling.
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l Meeting Summary The licensee indicated that the 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period for commencing valve testing could impose a hardship in those cases of " unscheduled" shutdowns where the personnel involved in valve testing would be the same individuals involved in corrective maintenance of the equipment which caused the shutdown.
The staff indicated that the licensee should commit to the staff definition for cold shutdown valve testing or provide an alternate definition for evaluation.
'The staff emphasized that the current review is limited to Class 1, 2 and 3 safety related valves (those that mitigate the consequences of an accident and/of safely shutdown the reactor and to maintain the reactor in a shutdown condition), the staff is not taking the prition that the other Code Class 1, 2 and 3 valves are not important and.therefore, should not be-tested.
It is possible that at some future tinie the staff f
could conceivably identify valves other than those safety related valves discussed in the working session that should be included in. the ISI program.
The staff indicated the need for an augmented inservice valve testing program for those valves which perform a pressure isolation function.
The staff position which was related to the licensee follows.
There are several safety systens connected to the reactor coolant pressure boundary that have design pressures that ar below the reactor coolant system operating pressure.
The staff has required that valves forming the interface between these high and low pressure systems are not subjected to pressuras which exceed their design limits.
In this role the vahes are performing a pressure isolation function.
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The redundancy provided by these valves regarding their pressure isolation function is important.
It is necessary to provide assurance that the condition of each of these valves is adequate to maintain system integrity.
For this reason, some methods, such as leak testing, should be used to assure their pressure isolation function.
In the event that leak testing is selected es the appropriate procedure fo'r reaching this objective, the staff believes that the following valves should be leak tested in accordance with IWV-3420 of Sect 4on XI of its applicable edition of the ASME Code:
RHR System Ell - F008, Ell - F009 Any 2 of:
Ell - F050A, Ell - F015A, Ell - F017A Any 2 of:
Ell - F050B, Ell - F0158, Ell - F017B HPCI System E41, F006 Core Spray F006A, F006B, F005A, F005B t
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Meeting Summary,
The licensee's revised submittal should address the pressure isolation valves and his method for satisfying the staff's requirements.
.If a revised (updated) inservice inspection or testing program conflicts with the Technical Specifications, the specification should be revised to conform them to the updated program.
Technical Specifications are considered to be "in conflict" only in cases where the requirements of the regulation (thus the requirements of the updated
' program) are more restrictive than the requirements of the Technical Specifications.
In such cases the licensee must propose changes to conform the Technical Specifications to the revised program.
In cases where the updated program is less restrictive than particular Technical Specification requirement, the, licensee must continue to comply with the Technical Specifications until he requests and is issued a Technical Specification change, The NRC staff will review such a proposed Technical Specification change to determine if it is acceptable or wh' ether the existing requirements should be retained as a augmented requirement pursuant to 50.55a(g){6)(ii).
3.
0_ pen Items for Staff Accion A list of valves were identified by the staff (Enclosure 3) which a.
are manual maintenance valves.
The staff asked how the ifcensee assures that these valves are in the proper position.
In repTy, it was stated that the valve line-up procedures assure proper valve position.
These procedures include the correct position for maintenance valves.
The staff requested that they be provided the line-up procedure number associated with each of these valves. This information will be provided.
The staff advised that this information does not need to be included as part of the program description.
Also, certain maintenance valves are provided with locks and chains for various non-safety reasons and the licensee did not feel that they should be categorized as E.
The staff questioned this point and will provide advice on their position, b.
The staff also questioned operability testing of the ADS safety relief valves.
They were advised that the SRV's do not receive a flow capacity test.
The staff identified this as an open item and will investigate further.
Meeting Summary c.
The licensee stated that the HPCI containment isolation valves cannot be isolated during cold shutdown unless the control circuits are jumped.
However, they can be tested on a quarterly basis during power operation without any difficulties.
The staff indicated that such testing may be in conflict with their guidelines on valve testing of redundant components. The staff will investigate this matter.
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Day d M. Ve/ elli, Project Manager Op ating Reactors Branch #3 Division of Operating Reactors
Enclosures:
1.
List of Attendees 2.
Staff Comments on ISI 3.
List of Maintenance Valves t
ENCLOSURE 1 NRC MEETING ON INSERVICE INSPECTION FEBRUARY 22, 1979 Name Organization V. Nerses NRC 5
D. Verrelli NRC T. McHenry NRC - Region II B. Rulenal-NRC Region II C. T. Moore Georgia Power Company J. A. Betsill Georgia Power Company R.' Baker Georgia Power Company H. Nix Georgia Power Company J. Watson Georgia Power Company T. Elton Georgia Power Company J. Edwards Georgia Power Company M. Kehoe Georgia Power Company T. Milton Southern Company Services 4 54/)
J. McLeod J A-Southern Company Services (
T. Caudie Southern Company Services M. Bel ford v4c e Southern Company Services ( tor). ?7 0 - 6V/2-J. King Southern Company Services G. Selby Battelle N.W. Labs T. Taylor Battelle N.W. Labs G. Lyon Battelle N.W. Labs s
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1 ENCLOSURE 2 Specific components needed for the following requests:
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1.
Request 2.1.3, specific valves which cannot be inspected to Code I
requirements and to which the justification applies.
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Requests 2.1.4, specific welds which are inaccessible for examination.
3.
Table 1, Item 3 4.5, Category B-J, circumferential and longitudinal welds which are scheduled to be examined this inspection period
. and which are characterized by the impracticalities given in NOTE 7 should be identified.
4.
Reques t 5.1.1,' no...e and/or number of the vertical centrifugal pumps included in the testing program fcr which relief is being requested should be given.
5.
Request 6.1.1, system name and valve number for which corrective action by Technical Specifications instead of the Code is being requested should be given or indicated in the submittal.
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ENCLOSURE 3 Category E* Valves - Normally opened or closed manual valves controlled
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by plant procedures.
1 Standby Liquid Control - F008, F001 A & B, F002A, F003A & B RHfi - F067, F060A & B, F034A - D, F002A & B, F014A & B Core Spray - F010A & B, F007A & B Nuclear Boiler System - F0ll A & B HPCI - F010 Chill Water System - F001, F002, F003, F005, F007A - D, F0ll., F008A - D, FC )6, F004, F012, F063, F016, F015 4
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