ML20115J824

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Rev 1 to Proposed Tech Spec Change 84-01,removing Details of ASME Section XI Inservice Insp Program
ML20115J824
Person / Time
Site: Pilgrim
Issue date: 04/17/1985
From:
BOSTON EDISON CO.
To:
Shared Package
ML20115J815 List:
References
NUDOCS 8504240120
Download: ML20115J824 (11)
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Text

{{#Wiki_filter:. LIMITING CONDITf0NS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.D Safety Relief Valves (Cen't) E. Jet Pumps 4 from the initial discovery of Whenever there is recirculation discharge pipe temperatures in flow with the reactor in the excess of 212*F for more than startup or run modes, jet pump 24 hours without prior NRC operability shall be checked daily approval of the engineering by verifying that the following evaluation delineated in 3.6.D.3. conditions do not occur simultaneously.

5. The limiting conditions of operation for the instrumentation 1. The two recirculation loops have a that monitors tall pipe tempera- flow imbalance of 15% or more when ture are given in Table 3.2.F. the pumps are operated at the same speed.

E. Jet Pumps

2. The indicated value of core flow
1. Whenever the reactor is in the rate varies from the value derived startup or run modes, all jet from loop flow measurements by more pumps shall be operable. If it than 10%.

is determined that a jet pump is inoperable, an orderly shutdown 3. The diffuser to lower plenum shall be initiated and the differential pressure reading on an reactor shall be in a Cold individual jet pump varies from Shutdown Condition within established jet pump P 24 hours. characteristics by more than 10%. F. Jet Pump Flow Mismatch F. Jet Pump Flow Mismatch

1. Whenever both recirculation Recirculation pump speeds shall be pumps are in operation, pump checked and logged at least once speeds shall be maintained within per day.

10% of each other when power level is greater than 80% and within 15% of each other when power level is less than or equal to 80%.

2. If Specification 3.6.F.1 is exceeded immediate corrective -

action shall be taken. If recirculation pump speed mismatch is not corrected within 30 minutes, an orderly shutdown shall be initiated and the reactor shall be in the Cold Shutdown condition within 24 hours unless the recirculation pump speed mismatch is brought within limits sooner. 8504240120 850417 PDR ADOCK 05000293 P pm Amendment No. 127

LIMITING CONDITIONS FOR OPERATf0N SURVEILLANCE REQUIREMENTS 3.6.G Structural Integrity 3.6.G Structural Integrity

1. The structural integrity of Inservice inspection of the primary system boundary components shall be performed shall be maintained at the in accordance with the PNPS level required by the ASME Inservice Inspection Program.

Boiler and Pressure Vessel The results obtained from Code, Section XI " Rules for compliance with this program Inservice Inspection of Nuclear will be evaluated at the Power Plant Components", completion of each ten year Articles IHA, IHB, IHC, IND and interval. The conclusions of IHF and mandatory appendices as this evaluation will be reviewed required by 10CFR50, Section with the NRC. 50.55a(g), except where specific relief has been granted by the NRC pursuant to 10CFR50, Section 50.55a(g)(6)(1). Amendment No. 127A

LIMITING CONDITIONS FOR OPERAT10N SURVEILLANCE REQUfREMENTS 3.6.H High Energy Piping (outside 4.6.H High Energy Piping (outside containment) containment)

1. The high energy line sections The inspections listed in identified in Table 4.6.2 shall Table 4.6.2 shall be performed be maintained free of visually as specified to verify the '

observable through-wall leaks. structural integrity of the specified high energy line

2. If a leak is detected by the sections. The visual I survelliance program of 4.6.H. Inspection for leakage shall efforts to identify the source be consistent with the of the leak shall be started requirements of ASME Boller and immediately. Pressure Vessel Code, Section XI, 1980 Edition, Winter 1980
3. If the source of leakage cannot Addenda, Subarticle IWA-5240.

be identified within eight hours of detection or if the leak is found to be from the pressure retaining boundary in the sections identified in fable 4.6.2, the leak shall be isolated or the reactor shall be in a cold shutdown condition within 48 hours.

4. When the modifications, described in FSAR Amendment No. 34, to provide protection against high energy line breaks outside of the primary containment have been completed, Technical Specifications 3.6.H and 4.6.H
      ~            will no longer be required.

Amendment No. 1278

S 9 e 9 PAGES 129 THROUGH 136 ARE DELETED 1 l l I i

r 

    .                                        TABLE 4.6.2 INSPECTION REQUIREMENTS FOR HIGH ENERGY                        l LINES OUTSIDE CONTAINMENT-ITEM NO.            HIGH ENERGY AREA         INSPECTION METHOD
  • FREQUENCY
1. Main steam lines outside Visual Monthly When containment from contain- Operating ment to turbine stop valves
2. HPCI steam line in torus Visual Monthly When area and in HPCI turbine Operating area
3. RCIC steam line in valve Visual Monthly When compartment and pump Operating compartment
4. RWCU line in pump, heat Visual Monthly When exchanger compartments and Operating valve compartment
5. Feedwater lines outside Visual Monthly When containment to the reactor Operating feedwater pump check valves A visual inspection for indications of leakage from all design basis piping break locations.

137 l 

     ~

LIMITING CONDITIONS FOR OPERATION SURVEZLLANCE REQUIREMENTS 3.6.I Shock Suppressors (Snubbers) 4.6.I Shock Suppressors (Snubbers)

1. During all modes of operation The following surveillance except Cold Shutdown and Refuel, requirements apply to all safety all safety-related snubbers related hydraulic and mechanical listed in PNPS Procedures shall snubbers listed in PNPS Procedures.

be operable except as noted in 3.6.I.2 through 3.6.I.3 below. The required visual inspection interval varies inversely with An Inoperable Snubber is a the observed cumulative number properly fabricated, installed of inoperable snubbers found and sized snubber which cannot during an inspection. pass its functional test. Inspections performed before that interval has elapsed may be Upon determination that a snubber used as a new reference point to is either improperly fabricated, determine the next inspection. Installed or sized, the However, the results of such corrective action will be as early inspections performed specified for an inoperable before the original time snubber in Section 3.6.I.2. interval has elapsed may not be used to lengthen the required

2. From and after the time that a interval.

snubber is determined to be inoperable, replace or repair Number of snubbers found the snubber during the next 72 inoperable during inspection or hours, and initiate an engineering during inspection interval: evaluation to determine if the components supported by the Subsequent snubber (s) were adversely affected Inoperable Visual Inspec-by the inoperability of the Snubbers tion Interval snubbers and to ensure that the supported component remains 0 18 Months i 25% capable of meeting its intended 1 12 Months i 25% function in the specific safety 2 6 Months i 25% system involved. 3,4 124 Days t 25% 5,6,7 62 Days 1 25% Further corrective action for 8 or more 31 Days 1 25% this snubber, and all generically susceptible snubbers, shall be The required inspection interval determined by an engineering shall not be lengthened more evaluation. than one step at a time.

3. From and after the time a snubber Snubbers may be categorized in is determined to be inoperable, two groups, " accessible" or improperly fabricated, improperly " inaccessible" based on their installed or improperly sized, if accessibility for inspection the requirements of Section(s) during reactor operation. These 3.6.I.1 and 3.6.I.2 cannot be met, two groups may be inspected then the affected safety system, independently according ts the or affected portions of that above schedule.

system, shall be declared inoperable, and the limiting 1. Visual Inspection Acceptance condition for that system Criteria entered, as appropriate. A. Visual inspections shall verify: Amendment No. 137a i

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS 3.6.I Shock Suppressors (Snubbers) 4.6.I Shock Suppressors (Snubbers)

4. Snubbers may-be added to, or 1. That there are no visible removed from, per 10 CFR 50.59, indications of damage or safety related systems without impaired operability, prior NRC approval. The addition or. deletion of snubbers shall be 2. Attachments to the reported to the NRC in accordance foundation or support with 10 CFR 50.59. structure are such that the functional capability of
the snubber is not suspect.

B. Snubbers which appear INOPERABLE as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval provided that:

1. The cause of the rejection is clearly established and remedied for that particular snubber, and
2. The affected snubber is functicnally tested, when i necessary, in the as found condition and determined OPERABLE per specifications i

4.6.I.2.B., 4.6.I.2.C., as applicable.

3. For any snubber determined inoperable per specification 4.6.I.2, clearly establish'the.cause of rejection and remedy the problem for that snubber, and any generically susceptible snubber.
2. Functional Tests (Hydraulic and Mechanical Snubbers)

A. Schedule At least once per operating cycle (18 months), a representative sample (10% of the total of each type: s hydraulic, mechanical) of snubbers in use in the plant shall be functionally , tested, either in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria of Amendment.No. 137b 

      ' LIMITING CONDITIONS FOR OPERATION   SURVEILLANCE REQUIREMENTS 4.6.I     Shock Suppressors (Snubbers) of each snubber, the date at which the designated service life commences and the installation and maintenance records on which the designated service life is based shall be maintained.

B. At least once per cycle,'the installation and maintenance records for each safety related snubber listed in PNPS Procedures shall be reviewed to l verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled snubber service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated, or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This reevaluation, replacement or 

                                          ~     reconditioning shall be indicated in the records.

C. This Snubber Service Life Monitoring Program shall become effective July 1, 1982. Amendment No. 137d

PAGES 137e THROUGH 1371 ARE DELETED s

m 9 Bases: 3.6.G and 4.6.G Structural Integrity The. Pilgrim Nuclear Power Station Inservice Inspection Program conforms to the requirements of 10 CFR 50, Section 50.55a(g). Where practical, the inspection of ASME Section XI Class 1, 2, and 3 components conforms to the edition and addenda of Section XI of the ASME Boller and Pressure Vessel Code required by 10 CFR 50, Section 50.55a(g). When implementation of an ASME Code required inspection has_been determined to_be impractical for PNPS, a request for relief from the inspection requirement is submitted to the NRC in accordance with 10 CFR 50, Section 50.55a(g)(5)(lii). Requests for relief from the ASME Code inspection requirements will be submitted to the NRC prior to the beginning of each 10 year inspection interval for which the inspection requirement is known to-be impractical. Requests for relief from inspection requirements which are identified to be impractical during the course of the inspection interval will be reported to the NRC on an annual basis throughout the inspection interval. As permittedlby 10 CFR 50, Sections 50.55a(g) and 50.55a(b)(2), the extent of examination (i.e. the number of welds to be examined and the welds selected for examination) has been modified for pressure retaining welds in ASME Code Class 1 and 2 piping as follows: ASME ASME Code Code Class Category Modification 1 B-J The requirements of Table IWB-2500 and Table IWB-2600 Category B-J of Section XI of the ! ASME Code, 1974 Edition, Summer 1975 Addenda are the basis for determining the extent of examination. I 2 C-F The requirements of paragraph IWC-1220, Table ' IWC-2520 Category C-F and C-G, and paragraph [ IWC-2411 of Section XI of the ASME Code, 1974 Edition, Summer 1975 Addenda are the basis for determining the extent of examination. I In addition to the requirements of 10 CFR 50, Section 50.55a(g), Boston Edison Company has established an augmented inservice inspection program for certain ASME Code Class 1, Category B-J pressure retaining welds designated as Group I welds. The Group I welds shall be inspected during each ten year interval to provide additional conservatism in the overall approach of protection against pipe whip which has the potential to breach the containment. 150' m

  • Bases:

3.6.H and 4.6.H High Energy Piping Outside of Containment Analyses performed and submitted to the AEC as Pilgrim Nuclear Power Station, Unit #1, FSAR Amendment #34 indicate that certain modifications to the station increase the protection against the potential effects of postulated high energy piping failures outside the primary containment. In order to provide greater assurance that the integrity of the high energy piping outside the primary containment is maintained at an acceptable level in the interim until these modifications can be completed, an increase in the frequency of inspections of the areas of concern has been initiated. The monthly visual inspection of high energy piping outside the containment while the station is operating provides greater assurance of the timely detection of postulated piping failures and allows appropriate corrective action to be performed. Reference to Subarticle IWA-5240 of the ASME Boiler and Pressure Vessel Code, Section XI, 1980 Edition, Winter 1980 Addenda, ensures that appropriate visual examination techniques are used to implement the requirements of Technical Specification Table 4.6.2. These visual examinations ~will normally be made with the indicated piping and insulation in its operating condition. Subsequent to the completion of the modifications, the inservice inspection requirements defined in Section 4.6.G of these Technical Specifications will provide adequate inspections to allow timely detection of postulated failures. Amendment No. 151 =}}

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