IR 05000277/1979022
| ML19257A229 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 11/09/1979 |
| From: | Caphton D, Foley T, Tanya Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19257A228 | List: |
| References | |
| 50-277-79-22, 50-278-79-24, NUDOCS 8001030110 | |
| Download: ML19257A229 (11) | |
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I 50-277/79-22 Report No. 50-278/79-24 50-277 Docket No. 50-278 DPR-44 License No. DPR-56 Priority
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Category C
Licensee:
Philadelphia Electric Company
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2301 Market Street Philadelphia, Pennsylvania 19101 Facility Name:
Peach Bottom Atomic Power Station, Units 2 and 3 Inspection at:
Delta, Pennsylvania Inspection conducted:
September 17-21, 1979 u/7/77 Inspectors:
a4 T.Fo]6, React (rInspector date si~gned
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hl N
11 f1 f 29
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T. H. SMth, Reactor Inspector date signed f-p; date signed Approved by:
1, &/
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D. L5 Ca'phtoli/,' Chief, Nuclear Support det6/ signed Section 1, RO&NS Branch Inspection Summary:
Inspection on September 17-21, 1979 (Combined Report Nos. 50-277/79-22 and 50-278/79-24)
Areas Inspected:
Routine, unannounced inspection by regional based inspectors of surveillance of pipe support and restraint systems; preparation for refueling; refueling activities; refueling outage related maintenance and surveillance; and licensee action on previous inspection findings.
The inspection involved 69.5 inspector-hours (Unit 2 - 21.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, Unit 3 - 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) on site by two NRC regional based inspectors.
Results:
No items of noncompliance were identified.
1666 100 Region I Form 12 (Rev. April 77)
8001030
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DETAILS 1.
Persons Contacted The below listed technical and supervisory personnel were contacted:
Philadelphia Electric Company
- J. Austin, Contruction Field Engineer
- W. Barley, Engineering Health Physic.s L. Griffithe, Bechtel Lead Coordinator T. Hinkle, ISI Coordinator
- C. Mengers, QA Site Supervisor
- J. Mitman, Test Engineer F. Polaski, Reactor Engineer S. Roberts, Results Enginee.r J. Spencer, Maintenance Engineer S. Spitko, Assistant Maintenance Engineer
- W. Ullrich, Station Superintendent
- A. Wasong, Test Engineer J. Yacyshyn, Test Engineer The inspectors also talked with and interviewed other members of the engineering and operating staffs.
- denotes those present at the exit interview.
2.
Licensee Action on Previous Inspection Findings (Closed) Unresolved Item (277/77-15-01):
The licensee currently has Procedure ST 9.15-2, " Seismic Hydraulic Snubber Inspection," Revision 9, dated September 21, 1979, which incorporates acceptance criteria for snubber piston settings to ensure that snubbers will have sufficient stroke available for thermal growth and identifies action to be taken in the event a snubber is found outside of the acceptance criteria.
Additionally, the revised procedure identifies the action to be taken when a snubber is found in the locked-up or frozen condition.
This item is closed.
(Closed) Unresolved Item (277/77-19-06):
The inspector reviewed Procedure M.65.4, Revision 0, " Hydraulic Snubber Testing", dated September 14, 1979, which combines and supersedes Procedures M18.1 and M18.4, which discuss maintenance and functional testing of snubbers.
The inspector noted that the new procedure sets and verifies the bleed 1666 101
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rate first then sets and verifies the lockup velocity.
This precludes the possibility of changing the lockup velocity of the snubber if a large change in bleed rate were made after setting the lockup rate.
This item is closed.
(Closed) Unresolved Item (278/78-32-02 and 277/78-29-01): The inspector reviewed Procedure ST 13.31, " Snubber Functional Test", Revision 1, change dated September 18, 1979 and a letter from E. Kister to T. W.
Ullrich dated September 18, 1979, as referenced in the above procedure.
The procedure now incorporates new acceptance criteria which accounts for the change in fluid viscosity and hence, the lockup rate at elevated temperatures, and requires that future snubber functional testing be
~ performed within a certain temperature band.
This item is closed.
(Closed) Unresolved Item (277/78-02-03 and 278/78-02-03):
The inspector reviewed Procedure ST/ISI 5, Inservice Inspection - Visual Examination of Hydraulic Restraints, Spring Hangers, Rigid Hangers, Rigid Supports and Anchors, and inspection results from 1977 and 1978 for Unit 2.
The inspector noted that the licensee's approved program included an exemption from verifying piston settings for hydraulic snubbers in accordance with Category B-K-2 of Section XI of the ASME Code.
This exemption was based upon the existence of the licensee's snubber Technical Specification (TS) surveillance and testing program.
Further review and discussions with the licensee's representative indicated that nonsafety related as well as safety related snubbers were included in the licensee's surveillance program.
Piston rod extension is required to be recorded and action is required if the measurement recorded is not within the provided acceptance criteria.
This satisfies the requirement of the ASME Section XI category B-K-2 relating to snubbers.
This item is closed.
3.
Su_rveillance of Pipe Support and Restraint Systems a.
General The inspector reviewed general surveillance procedures for hydraulic snubber visual inspection and functional testing.
The inspector observed the maintenance crew during the reassembly of several hydraulic snubbers and also observed the setpoint adjustments and testing of one hydraulic snubber.
The inspector toured the Drywell, Torus Room, Residual Heat Removal (RHR) Pump Rooms, High Pressure Coolant Injection (HPCI) Pump Room and the Reactor Core Isolation Cooling (RCIC) Pump Room.
The inspector observed the general conditions of support components including hangers, clamps, braces, brackets, turnbuckles, lugs, clevis and support base plates, and visually verifed the following:
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No evidence of corrosion, cracks or other detrimental indica-
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tions.
No observable deformation.
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Nuts, bolts and fasteners were secure.
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Support settings were within the designated bands.
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Chafing with other components was not observed to result
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from pipe mo/ ament due to thermal growth.
b.
Observations (1)
During this inspection, the inspector identified two snubbers 23 HBS-37 and 23 DBN-S-27 with piston rod extensions, one at the fully retracted position and the other near the fully extended position, thus potentially over stressing the pipe to which they are attached.
This matter is unresolved pending the resetting of these snubbers' piston extension and an evaluation of the pipe stress level if it is determined that the snubbers did fully retract or extend during thermal growth (277/79-22-01).
(2) Additionally, the inspector verified design measurements on numerous support components by comparing actual dimensions to those required by design drawings provided by the licensee.
The following supports ware noted to be not specifically in accordance with the design drawing:
(a) Support 3-23-DDN-S-29 on ITT Grinnell sketch 3-23-45 or HISO-2355-A.
Extension piece, item 2 was measured to be 35 inches vice the required 3 feet, 2 inches.
(b)
Support 3-13-HB-S-23 on Grinnell HISO 1354 or sketch 3-13-32.
Detail
"A" Baseplate bolt to bolt dimension was measured to be 8 inches vice the required 7 inches and was missing one nut from the four required anchor bolts.
This nut was replaced prior to the exit interview and the licensee's representative stated that the bolt was inadvertently left off duririg anchor bolt testing, however, the Quality Control checks had not been completed at the time of identification.
(c) Support 3-23-DDN-S23A on Grinnell sketch 3-23-38A or HISO 2355A.
The end weld of the angled I-beam to the end of the horizontal I-beam measure 21/2 inches vice the required 8 inches.
This discrepancy was noted 1666 103
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under the licensee's inspection program in response to IE Bulletin 79-14 and is presently being evaluated.
The above items are collectively unresolved pending NRC review of the licensee's evaluation of each of the above discrepancies (278/79-24-01).
c.
Review of Program and Procedures (1) The inspector reviewed the following documents to ascertain whether the licensee's program was approved and in accordance with Technical Specifications and licensee commitments.
(a)
Procedure M 65.1, " Hydraulic Snubber Overhaul", Revision 0, September 14, 1979.
(b) Procedure M 65.4, " Hydraulic Snubber Testing", Revision 0, September 14, 1979.
(c) Procedure ST 9.15-2, " Seismic Hydraulic Snubber Inspec-tion Unit 3 Only", Revision 8, September 4, 1979.
(d) Procedure ST 9.15-1, " Seismic Hydraulic Snubber Inspec-tion Unit 2 Only", Revision 4, October 26, 1977.
(e)
Philadelphia Electric Company letter from E. Kister, Mechanical Engineering Division to W. Ullrich, Station Superintendent, dated January 3,1979.
(f)
Philadelphia Electric Company letter from E. Kister to W. Ullrich, dated September 18, 1979.
(g) Procedure M 18.8, " Hydraulic Snubber Operability Check",
Revision 1, February 10, 1979.
(h) Procedure M 18.4, " Hydraulic Snubber Lockup and Bleed Verification", Revision 5, January 11, 1979.
(i)
1979 Inservice Inspection Report for Peach Bottom Atomic Power Station Unit (3) January 1979.
(j) Peach Bottom Atomic Power Station Inservice Inspection Program for Unit 2 and Common Plant.
(k)
Procedure ST/ISI-5, " Inservice Inspection - Visual Examination of Hydraulic Restraints, Spring Hangers, Rigid Supports and Anchors", Revision 2, September 26, 1978.
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(1) Energy Conversion Research Section 510-1, letter from E. Kohler to C. Mengers, dated September 14, 1979.
(m) Procedure ST 13.31, " Snubber Functional Test", Revision 1, October 26, 1977.
(2) With the exception of the items listed below, the inspector identified no significant problems.
(a) The licensee's functional test procedure ST 13.31 requires that snubbers be tested in accordance with Procedure M.65.4, " Hydraulic Snubber Testing", which performs a lockup test (6) times and a bleed rate test (2) times on each snubber tested.
The procedure does not specify the following:
(1) Which data point is to be compared with the provided acceptance criteria?
(2) What action is to be taken when a data point varies significantly from the mean of the data?
(3) What is a significant deviation from the mean?
The licensee's representative stated that the procedure would be revised to incorporate statements addressing the above concerns prior to the upcoming Unit (2) refueling outage.
This matter is unresolved (277/79-22-02 and 278/79-24-02).
(b) The licensee's latest completed and reviewed snubber surveillance procedures ST 9.15-2, for accessible and inaccessible snubbers were completed on October 5, 1978 and May 16, 1978.
Procedure ST 9.15-2 completed June 5, 1979 had not been formally reviewed at the time of this inspection three months after its completion.
However, a review of Maintenance Request Forms (MRF's)
and discussions with the snubber test engineer revealed that action had been taken correcting deficiencies noted on the surveillance procedure.
The licensee's present snubber surveillance procedura inspects snubbers whenever practicable in addition to the TS requirements.
The present TS inspection interval requirement for Unit 3 is 18 months, which had not lapsed since the May and June surveillance inspection.
The liceraee has recently 1666 105
completed a surveillance inspection of all Unit 3 snubbers during the present outage and had not completed the formal review of these procedures prior to the exit interview.
This matter is unresolved pending a timely review of the required surveillance which is used to satisfy thi. TS surveillance interval (278/79-24-03).
d.
Snubber Orifice Modification The licensee currently has a program established to either enlarge or remove the fixed bleed orifice from hydraulic snubbers.
The licensee believes this is advantageous because the bleed orifice is only required to maintain bleed rates of approximately 1/8"/ min and is not required for settings of 1/2"/ min and larger, and reduces the possibility of the orifice becoming blocked by contami-nants.
The inspector reviewed the Safety Evaluation correspondence with ITT Grinnell Corporation and results of tests conducted by Peach Bottom Atomic Power Station relating to the snubber modification which removed the bleed orifice from 21/2" hydraulic snubbers.
The Safety Evaluation documented in PORC minutes 78-82-2, stated that this modification is recommended by the manufacturer and has been independently reviewed and determined that it does not involve an unreviewed safety question or a change to the TS.
The inspector had no further questions on this item.
e.
Additional Snubber Monitoring The inspector noted that the licensee has initiated a program, in addition to Technical Specification requirements, that requires monitoring of snubbers, additional testing, rebuilding, sampling and analyzing hydraulic fluid, evaluating faulted snubbers prior to rebuilding and correlating snubber serial number, location and environment to snubber problems, and evaluating the history of each snubber.
Additionally, for added conservatism, the 10 snubbers selected for TS related functional testing were selected after the TS required visual surveillance and the selection was based upon the appearance of possible off-normal conditions.
The inspector had no further questions in this area.
f.
Inservice Inspection of Class 1 and 2 Components 1666 106
Section XI of the ASME Code, Examination Category B-K-2 and C-E-2 calls for the verification of support settings of constant and variable spring type hangers, snubbers and shcck absorbers.
At present, the licensee appears to be meeting the requirements of Category B-K-2 for class 1 supports, however, no evidence exists of inspecting class 2 supports at this time.
The licensee's program for inservice inspection does require that class 2 components be inspected and one third of all class 2 supports must be inspected prior to the end of the present 40 month inspection interval.
This item is unresolved (277/79-22-03 and 278/79-24-04).
4.
Refueling Preparations a.
Documents Reviewed The inspector reviewed the following documents to verify that the licensee had approved, and technically acceptable procedures for the present refueling outage.
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FH-6C, " Fuel Movement and Core Alteration Procedure During a Fuel Handling Outage", Revision 8, dated September 18, 1979.
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FH-5, "New Fuel Inspection, Channeling and Placement in the Fuel Pool", Revision 16, dated September 11, 1979.
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FH-26.1, " Pressure Sipping", Revision 2, dated April 6, 1979.
b.
Fuel Receipt and Inspection Procedures and check off lists related to the receipt and inspection of 272 fuel assemblies and fuel channels were reviewed.
Licensee fuel inspector qualification records were also checked.
No discrepancies were identified and the inspector had no further questions in this area at this time.
c.
Fuel Handling The inspector reviewed the fuel handling procedure, FH-6C, and verified that refueling related Technical Specification requirements were properly addressed.
No discrepancies were identified and the inspector had no further questions in this area at this time.
5.
Refueling Activities a.
General
The inspector verified by direct observation and record review that fuel handling activities were being conducted in accordance with approved procedures and Technical Specification requirements.
Items inspected included:
(1) Pre-fuel handling surveillance testing; (2) Plant staffing, including that ', che control room and in the refueling area, was as specified in the Technical Specifi-cations; (3) Refueling area housekeeping was satisfactory; (4) Communications were maintained between the refueling bridge and the control room; (5) Vessel water level and clarity were acceptable; (6) Fuel insertion was conducted in accordance with procedures; (7) Source range monitoring was performed per the Technical Specifications; (8) Radiation monitoring was acceptable; (9) Secondary containment integrity was maintained; (10) A senior Reactor Operator with no concurrent duties directly supervised the fuel handling; and, (11) Fuel accountability was as specified.
No discrepancies were identified.
b.
Test Witness The inspector witnessed the performance of the Source Range Monitor (SRM) daily response check of Procedure ST 3.1.2, "SRM Core Monitoring Test", Revision 3, conducted on September 21, 1979.
No discrepancies were identified.
The inspector had no further questions on the refueling at this time.
6.
Surveillance 1666 108
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a.
Scope The inspector reviewed the documentation related to the local leak rate testing of the Main Steam Isolation Valves (MSIV)
during the current refueling outage.
The documents were reviewed for completeness, technical adequacy and acceptability of the results based on Technical Specification and 10 CFR 50, Appendix J requirements.
b.
Documents Reviewed ST 30.021, " Inboard MSIV LLRT A0-3-2-80A (B, C, D)", Revision
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4, dated September 4, 1979.
ST 30.022, " Outboard MSIV LLRT A0-3-2-86A (8, C, 0)", Revision
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1, dated September 4, 1979.
c.
Findings The inspector noted in his review that seven of eight MSIV's failed to meet the Technical Specification leakage limit, and that the leakage of six of the MSIV's was in excess of the maximum ability of the test equipment to measure.
The licensee had not planned to report the results of these tests,.however, after discussions with the inspector, a representative of the licensee committed to report these test failures, as well as any additional local leak rate test failures, at the completion of the present outage.
The following items concerning the MSIV local leak rate test procedures are unresolved and are collectively designated item number 278/79-24-05.
(1) One section of test procedure ST 30.021 calls for the local leak rate test to be run for 15 minutes while another section calls for one hour.
This item is unresolved pending correction of the inconsistent time requirement.
(2)
Procedure ST 30.021 contains a rotameter correctiLn factor which does not appear to be understood by alI personnel conducting the tests.
In completed tests reviewed by the inspector, this factor was not consistently applied from test to test.
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(3) Using procedure ST 30.022, MSIV 86C was tested using a pressure decay test on the volume betwten the MSIV's with the reactor vessel also pressurized to test pressure.
Leakage of the inboard MSIV into the test volume would invalidate the pressure decay test results.
This item is unresolved pending procedural revision.
7.
Maintenance Activities a.
Documents Reviewed m
M 3.1, " Control Rod Drive Replacement", dated Septembrr 14,
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1979.
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M 3.4, " Control Rod Drive Repair", Revision 5, dated September 13, 1979.
M 1.3, " Main Steam Isolation Valves Maintenance", Revision
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5, dated August 17, 1978.
M 1.8, "MSIV Air Cylinder Repair", Revision 1, dated December
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1978.
b.
Maintenance Witnessed The inspector vitnessed the work being performed to repair the Main Steam Isolation Valves and verified that the work was being conducted in accordance with the above referenced approved proce-dures.
No discrepancies were identified.
8.
Unresolved Items Unresolved items are those items for which further information is required to determine whether they are acceptable or items of noncom-pliance.
Unresolved items are contained in Paragraph 3 and 6 of this report.
9.
Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1) at the conclusion of the inspection on September 21, 1979, and summarized the scope and findings of the inspection as they are detailed in this report.
During this meeting, the unresolved items were identified.
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