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U. S. ATOMIC ENERGY COMMISSION
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REGION I
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DIVISION OF COMPLIANCE Report of Inspection CO Report No. 219/70-5 Licensee:
JERSEY CENTRAL POWER AND LIGHT COMPANY Oyster Creek 1 License No. DPR-16 I
Category C Dates of Inspection:
May 18 - 22, 1970 Dates of Previous Inspection: April 21 - 23, 1970 (Special)
Inspected by:
R. J. Mc De rmo tt, Reactor Inspector (Responsible)
Date D. L. Caphton, Reactor Inspector Date Reviewed by:
R.'T. Carlson, Senior Reactor Inspector Date Proprietary Information:
None SCOPE Type of Facility:
Boiling Water Reactor Power Level:
1600 Mwt-Location:
Forked River, New Jersey Type of Inspection:
Routine, Announced Accompanying Personnel:
Mr. D. Sullivan, DRS (May 19, 1970)
Mr. D. Pomeroy, TSB, CO:HQ (May 20-21, 1970)
Mr. J. Tillou, CO:I (May 22,1970)
Scope of Inspection:
A meeting was held with management representatives in the Parsippany, N. J. offices ^ of JC on May 18, 1970, to discuss action taken on the part of management in response to the Commissioner Thompson - Regulatory - JC meeting held on March 25, 1970. The site was visited on May 19-21, 1970 to review plant operations since the last routine inspection.
Mr. D. Sullivan assisted the assigned inspector in reviewing protective circuitry relative to the primary system instrument line break discussed in Section F.2.
Mr. Tillou assisted the assigned inspector in reviewing matters relating to the core spray nozzle safe ends discussed in Section E.2.
julis was done at the Parsippany, N. J. engineering offices of GPU on May 22, 1970.
9500070212 950227 PDR FOIA DEKOK95-36 PDR
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SUMMARY
Safety Items - None g
Noncompliance Items - Six items of noncompliance were identified during the i
inspection:
1 1.
Contrary to the intent of Paragraph 3.1.A and Table 3.1.1, Item B.2'or 3 l
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of the technical specifications, the reactor was operated on January 1,
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1970 with a portion of the major steam break protection instrumentation being incapable of performing its intended function.
(See Section F.1.)
2.
Contrary to Paragraph 3.C. (1) of the Provisional Operating License, prompt notification and a timely written report was not provided the Commission of the partial loss of major steam break protective instrumentation discussed above in item 1.
(See Section F.1.)
3.
Contrary to Section 2.3.4) of the technical specifications, the reactor was operated on March 25, 1970 during a period when two (of four) relief valves would not have relieved automatically on a high reactor pressure of 51125 psig.
(See Section F.2.)
4.
Contrary to Paragraph 3.C.(1) of the Provisional Operating License, prompt notification and a timely written report on the loss-of relief protection discussed in item 3.above was not provided to the Consnission.
(See Section F.2.)
i 5.
Contrary to Section 4.2.D. of the technical specifications, during reactor operation on April 15, 1970, the control rods were not exercised daily during a period when there were two inoperative control rods.
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(See Section F.3.)
6.
Contrary to Paragraph 3.C.(1) of the Provisional Operating License, the written report on the failure of the main steam isolation valves to meet specified leakage limits during testing on April 21, 1970, was not provided to the Commission in a timely manner.
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Unusual Occurrences -
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1.
Main Steam Isolation Valve closure initiated a scram (No. 42) when system pressure decreased to the 850 psig setpoint.- The pressure was 5
allowed to reduce to this setpoint due to a sluggish turbine steam pressure controller.
(See Section C, Scram No. 42) 2.-
Primary system pressure instability resulted during a transfer from the mechanical pressure regulator to the electrical pressure regulator which are part of the turbine steam pressure controller. The instability resulted in a low reactor water level scram.
(See Section C, Scram No.
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Problems were experienced with the normal withdrawal of the control rods and generalized increased seal leakage was noted. The reactor was shut down for ufive weeks to inspect 137 (of 138 installed) control rod drives. Most of the drives were observed to have broken seals and 1
bulged index tubes.
(See Section F.3.)
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4.
Two (of four) Main Steam Isolation Valves were determined to be leaking in excess of. limits during testing on April 21, 1970. To date, the scali'g performance of these valves has been totally unsatisfactory. A n
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succeksful leak rate measurement test of all four valves has yet to be j
accomplished without prior maintenance.
(See Section K.1.)
i Closur\\ ' of an excess flow check valve in an instrumentation sensing 5.
e during! operation on January 1, 1970 resulted in a loss of major steam break protective and violated a Limiting Condition for Operation.
(See 1
Section F.1. and Noncompliance Item 1.)
6.
The breaking of a primary system instrument sensing line (and subsequent
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closure of an excess, flow check valve) on March 25, 1970, resulted in a loss of redundancy of protective instrumentation and also violated a Limiting Safety System Setting for primary system pressure relief protection.
(See Section F.2. and Noncompliance item 3.)
Status of Previously Reported Problems - None Other Significant Items -
1.
The corporate reorganization that resulted from the Commissioner Thompson -
Regulatory - JC meeting on March 25, 1970 (meeting minutes attached) were reviewed with Messrs. Sims and Verrochi, JC Vice Presidents on May 18, 1970.
Mr. Sims is now the responsible vice president for the operation of OC-1 (replacing Mr. Logan) and the responsibilities of the Manager of Generating Stations have been divided.
Mr. I. Finfrock, formerly Technical Supervisor OC-1, has been selected for the newly created position of Manager of Nuclear Generating Stations, JC.
(See Section A.1.)
2 A meeting was held with Mr. liirst, Chairman of the General Office Review Board (GORB) to discuss discerned weaknesses in the CORB's performance to date.
(See Section A.3.)
3.
Meetings were held with Mr. McCluskey, Station Superintendent, OC-1, to discuss CO concerns for the Plant Operating Review Committee's (PORC's) performance, expand on items requiring review by PORC as defined in the technical specifications, and to discuss CO:I expectations for PORC's future performance.
(See Section A.4.)
4.
One shift foreman terminated, leaving three shift foremen.
Mr. J. Carroll, Operations Supervisor (Acting) now must provide periodic coverage for this position. Three site people are scheduled to take SOP exams on July 29, 1970.
(See Section A.2.)
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3 4-5.
Snubbers (instrument pulse dampers) were disclosed to be installed in instrument lines in-at least two applications which provide plant safety functions. CO:I has requested GORB to' review the acceptability of these.
'i (See Section F.5.)
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6.
The scram circuit modifications that were previously made at Nine Mile
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Point are still under consideration by JC.
(See Section F.4.)
f 7.
Stack monitor sampling equipment has not been installed to determine the sampling line loss for the permanent stack sampler. JC does intend to install this and has stated that it will be installed and operational by August, 1970 8.
Linear defects were observed during a PT check of the north core spray nozzle safe.end overlay cladding. The investigation of the defects (boat sample) disclosed that the material was Inconel and not 308-L as stated in Amendment 43 and that the defects (microfissuring) were the result of weld solidification during the application of the cladding.
The licensee has removed all defects by grinding and has measured the remaining safe end wall thickness by radiographic techniques. JC considers that the measured wall thickness meets code requirements. JC also stated that Amendment 43 will be corrected. CO:I has advised JC that it would be prudent to perform additional wall thickness measurements in light of the indicated small margin over minimum code requirements.
(See Section E. 2.
Management Interview - Inspectors Caphton and McDermott conducted an exit interview with Messrs. McCluskey, Ross, Hetrick, Carroll, and Riggle on May 21, 1970 The following was discussed:
1.
Noncompliance Items a.
Rod Drive Exercising - The inspectors stated that there were no records that the control rods had been exercised on April 15, 1970 as required by technical specifications.when there are two inoperative rods.
Mr. McCluskey acknowledged this as an item of noncompliance and stated that the plant had just recovered from a scram on April 14, 1970 and that during the busy schedule this item was overlooked.
b.
Violating a Limiting Condition for Operation - The inspectors commented that a review of records and logs had indicated that major steam break protective instrumentation was removed in part during an
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operating period on January 1, 1970, when an excess flow check valve was closed. This was identified as an item of noncompliance. The
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inspectors also stated that the recorded sequence of events that
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caused scram No. 31 on December 31, 1970.which was documented in the g
scram report and operating logs, implied that the cause of this i'
prior to resuming operations on that date.
Mr. McCluskey did not scram (i.e., the excess flow check valve closure) was not determined cousnent other than to state he would review this matter.
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Violating a Limiting Safety System Setting - The inspectors stated that their review of the March 25, 1970 instrument line break had disclosed that the break resulted in a loss of automatic relief ability for two (of four) relief valves.
This was identified as an item of noncompliance.
Mr. McCluskey acknowledged this as an-item of noncompliance.
d.
Yailure to Report MS1V Leakage - The inspectors informed Mr. McCluskey-that the failure to provide a written report within 10 days of the j
f,ailure of the MSIV's.to meet leakage limits was an item of noncompliance Mr. McCluskey stated that JC had intended to cubmit a report, but cSnsidered that the report should be issued after repairs and final testing of the valves.
He acknowledged this as an item of non-compliance.
2.
Main Steam Isolation Valve Testing The inspectors stated that the performance of these valves to meet leakage requirements has been unsatisfactory to date. They further stated that more frequent testing of these valves should be considered in order to establish a level of confidence.
Mr. McCluskey stated that JC has previously committed to testing these valves during each plant cooldown until the next refueling outage. The inspectors again stated that testing in addition to this should be considered in view of the most recent failure of two (of four) valves to meet leakage limits.
No commitment was obtained from Mr. McCluskey to perform additional checks and it was inferred from the discussion that none would be performed.-
'3.
Stack Sampler,
Mr. McCluskey was informed JC has been extremely tardy in meeting prior commitments to install stack sampling equipment to measure line loss of the permanent stack sampler.
Mr. McCluskey assured the inspectors that the sampling equipment would be installed and operational by August, 1970 and that cold and wind conditions had precluded earlier installation.
4.
Snubberc Mr. McCluskey was asked if PORC had reviewed and approved the use of snubbers in instrument sensing lines.
He informed the inspectors it had.
Subsequent to the inspection Mr. McCluskey was requested during a telecon to recommend that GORB review this item'.
He informed the inspecter that GORB would be informed of' this request.
5.
North Core Spray Nozzle Mr. McCluskey informed the inspectors that JC considered Inconel overlay cladding on this and two other safe ends as acceptab_le and that the information presented in Amendment 43 stating that 308-L had been used, would be modified.
Subsequent to the inspection Mr. McCluskey was advised that a more prudent determination,should be made of the remaining wall thickness on the north core spray nozzle safe end.
Mr. McCluskey stated that this would be considered.
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6.
PORC Performance (See Other Significant Items 3 and Section A.2.).
1 DETAILS A.
Persons Contacted:
b Mr.
R. Sims, Vice President, Production Department, JC Mr. I. Finfrock, Jr., Manager, Nuclear Generating Stations, JC Mr. T. McCluskey, Station Superintendent, DC-1 Mr. D. Hetrick, Operations Supervisor, OC-1 Mr. J. Carroll, Operations Supervisor (Acting), OC-1 Mr. D. Ross, Technical Supervisor, OC-1 Mr. W. Riggle, Maintenance Supervisor, OC-1 Mr. W. Verrochi, Vice President, Design and Construction, GPU Mr. W. Hirst, Chairman, General Office Review Board, GPU Mr. B. Avers, Quality Assurance Manager, GPU Mr. B. Moore, Mechanica1 ' Engineer, GPU Mr. N. Goodenough, QA Engineer (Radiography), GPU Mr. D. Kaulback, Radiation Protection Supervisor, OC-1 B.
Administration and Organization 1.
Corporate Organization A meeting was held in Commissioner Thompson's office with Mr. R. F.
Bovier, JC President on March 25, 1970, to air Regulatory concerns with.
JC's top management regarding management support to and the conduct of the operation of the JC Dyster Creek 1 (OC-1) nuclear power station.
(See meeting minutes, Addendum 1 attached).
Discussions also included discerned weaknesses in the JC management system for 0C-1.
During this meeting Dr. Hann reviewed Regulatory's areas of concern:
lack of managerial attention; organizational weaknesses, particularly of technical personnel; loss of key personnel both at the site and at the corporate office level; and abnormal time required for the replacement of the personnel. Also discussed was the lack of timely reporting by JC to meet Regulatory requirements.
Messrs. Caphton and McDermott held discussions with Messrs. Hirst, Finfrock, Verrochi, and Sims on May 18, 1970 and the following significant organizational changes were reported.to have been made effective April 23, 1970:
a.
Mr. Sims has replaced Mr. Logan as the Vice President of the Production Department.
Mr. Sims reports to Mr. R. F. Bovier, JC President.
Mr. Logan was reported to have elected the option of early retirement.
b.
Mr. G. Ritter, Vice President, JC, is currently on an extended 1 eave of absence.
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The functions of the Manager of Production, JC,.have been divided f
into nuclear and non-nuclear categories.
Mr. I. Finfrock was
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assigned the duties of the Manager of Nuclear Facilities.
l Mr. Kelcec will continue on as Manager of Production but limiting i
his attention to only the non-nuclear JC plantr.
Mr. Finfrock
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previously held the position of technical supervisor at the DC-1 plant. Both Finfrock and Kelcec report directly to Sims. Mr.
Finfrock will also be responsible for the proposed Forked River 1 reactor and Three Mile Island Nos.1 and 2 (JC part owner).
d.
Mr. Sims infonned the inspector that Mr. Finfrock is responsible for all aspects of the OC-1 plant.
Mr. Finfrock and Mr. Sims reported that they now jointly review the OC-1 plant status on a daily basis.
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Messrs. Sims, Finfrock, and Hirst were questioned regarding the present JC recruitment program for site technical staffing.
Mr. Sims informed the inspectors that following the Commissioner Thompson - President Bovier meeting, Mr. Bovier met with Messrs. Finfrock and Sims and with the JC Vice President for Personnel.
Mr. Sims reported that although no additional personnel had been hired to date, an active, vigorous recruitment program is underway and some candidates were under current review.
He also reported that the two assistant technical engineers (J. Sullivan and E. Crowney) have now returned to the OC-1 site from the CE technical training course (Fuel Flux Management Course) in San Jose, California.
Mr. Sims also stated that Mr. McCluskey has Pickard and Lowe and Associates (consultants to JC) capabilities to call upon if needed for technical support.
Mr. Verrochi informed the inspectors that the Design and Construction Division of the CPU Service Company presently totals 65 technical people.
He projected that the size of this division would increase to 4 100 by j
1970 end.
Mr. Verrochi assured the inspectors that these personnel are available for technical support of the OC-1 plant, if and when requested.
He was asked who established priorities for personnel in the division as these personnel have responsibilities for many plants.
He stated that both Mr. Finfrock and Mr. Hirst are well aware of the significance of problems as they arise at OC-1, and that these personnel can be expected to establish the necessary priorities.
Mr. Verrochi also stated that Mr. Don Rees will' continue on for a time in the capacity of a full time OC-1 Project Manager, even though construction on the plant has been completed.
l 2.
Site Staffing Mr. McCluskey informed the inspectors that one shift foreman (Roth) recently quit after giving a very short notice. This currently leaves only three shif t foremen at the OC-1 site and requires Mr. J. Carroll, Operations Supervisor (Acting), to periodically provide relief for this position.
Mr. McCluskey also informed the inspectors that he currently
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i has plans to upgrade two licensed operators to the job of shift foreman and that three people were currently preparing to take the SOP exams (estimated to be at the end of July; 1970).
All GE technical support personnel have now lefr the site, l
3.
General Office Review Board (GORB) i L
The adequacy of GORB's past involvement in the audit of plant operations j
.p was discussed with'Mr. Hirst and Mr. Sims on May 18, 1970.
Mr. Hirst J
was questioned relative to GORB's source of information to enabic members to identify plant problems for further review.
Mr. Hirst stated that he, as chairman of GORB, relied for the most part on PORC minutes to identify problems at the OC-1 plant.
He also stated that Mr. Don Rees, Project Manager, OC-1, keeps him informed of problem areas.
Mr. Hirst was questioned by the inspectors how he disseminated information to other members.
Mr. Hirst stated that there is no systematic method at present to keep the other members fully informed of all the problems, but that he informs them as necessary. The inspectors informed Mr. Hirst that their
'l review of PORC meeting minutes had disclosed an almost total absence of CORB members attending these meetings. The technical specifications require the membership of PORC to include two members of GORB.
H'e informed the inspectors that better attendance by GORB members at future PORC meetings could be expected.
Mr. Hirst was encouraged by the inspectors to develop a systematic method for both obtaining information relative to operating problems at the plant and the distribution of same to the other members of GORB.
Mr. Hirst informed the inspectors that the continued involvement of Messrs. Heward and Ritter as members of the GORB is questionable at this time. No replacements for these two members have been chosen at the present time.
Resumes of all GORB members were supplied by Mr. Hirst and are retained in the Region I files.
4 Plant Operating Review Committee (PORC)
The adequt.cy of PORC's involvement in past plant' operations was discussed i
during the exit interview (see Other Significant Items, No. '3).
Mr. McCluskey was appraised of the inspectors' findings regarding the General Office Review Board's reliance on PORC meeting minutes to identify plant problems. The inspectors stated during this discussion that the PORC meeting minutes should receive added emphasis to ensure that PORC's review of operating problems at 0C-1 are adequately documented. The inspectors also commented that their review of PORC minutes during prior inspections had indicated a very marginal coverage of plant problems and in some instances no review had been made or, if made, had not been documented. The items requiring review by PORC that are specified in
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technical specifications were di.scussed in detail and expanded on for
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clarification to Mr. McCluskey.
Mr. McCluskey, at the conclusion of this discussion, stated that PORC has been slow in adequately documenting
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He also indicated.that some additional insight of PORC's responsibilities 2
was derived from the~ discussion.
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Licensing Activities y
A summary of correspondence, including licensee reports, issued in J
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conjunction with this facility since the last routine inspection I
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(See Addendum 2).
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Operations The reactor was shutdown on April 19, 1970 for a five week outage to inspect and repair the control rod drives.
(See Section F.2.) 'During this. outage the main steam isolation valves (MSIV's) were tested and two (of four).were determined to be leaking in excess of technical specification requirements.
(Discussed in Section K.)
Tabulated below are the scrams which have occurred since the last-routine inspection:
Date Cause April 7, 1970 Automatic scram from 100% power from the closure of (No. 4 )
the main steam line isolation valves. 'The Relay Department had requested relay trip checks on the spare exciter breaker.
During the relay testing, a direct short circuit was placed across the 125 V DC bus. This resulted in the dropout of other relays fed from this bus including three recirculation MG set relays, used d
for sensing loss of generator field. This in turn resulted in tripping three MG set drive motors and three recirculation pumps. Normally, reactor power and generator load would have been reduced but no scram would have occurred.
In this instance when the three recirculation pumps stopped, the reactor pressure 3
dropped to 850 psig before the generator steam control valves went fully closed. Upon reaching 850 psig, the main steam isolation valves' shut as required and
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initiated a reactor scram.
Investigation'of the un-
-expected pressure decrease disclosed that it took 12 seconds to stroke the steam electric pressure regulator s
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(EPR) and steam admission control valves from open to close. The normal. time should be approximately four 3
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The EPR was then stroked separately, and its time was found to be 6-1/2 seconds.
Further inspection g disclosed that a shock absorber mounted on a torque tube
- in the control linkages was found to be restricting the movement slightly. This was disconnected and adjusted, the EPR adjusted, and the total control system (including the steam admission control valves) was stoked in four seconds. The cause of the EPR time
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Date Cause discrepancy is related'to a failure to re-time this device after changeout of oil filters. GE-LSTG t
Division was notified of the problem and they will forward their recommendations to JC in the near future.
Three control rods (14-35,18-35 and 42-27) settled at the 02 position on this scram.*
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April 14, 1970 Automatic scram from a power level of 100% from a (No. 43) low reactor water level trip.
Scram was caused by.
steam pressure oscillations following an attempt to transfer steam pressure control from the mechanical pressure regulator (MPR) back to the electric pressure regulator (EPR), Control had been transferred to the MPR as part of a weekly test to verify its operability.
The steam pressure oscillations progressed to the point where voids were collapsed and resulted in a low reactor water level scram. Oscillations were the result of the operator raising the EPR pressure setpoint too rapidly when attempting to regain control with the EPR. A GE '
turbine representative was called in to check the EPR and found it to be in good working order.
It was his opinion that the regulator was brought in slightly too fast causing an overshoot which started the oscillations.
Instability of the linkage from the pressure regulator to the control and bypass valves is being investigated by GE at this time and modifications may be made at a 1ater date. All operators have been cautioned about i
the need to bring the EPR in very slowly as outlined in the turbine procedure.
If oscillations start again, the operators have been instructed to turn the power off the EPR which will transfer control to the MPR. At the time of the scram, rod 18-35.did not insert to 00 but inserted to 01 and latched at notch 02.**
This drive was subsequently scrammed four times with the accumulator charged and the charging. valve closed - twice with reactor pressure at 1000 psig and twice at 920 psig. This drive was also scrammed once with reactor pressure at 920 psig. During all scrams, the drive was observed to stop at the 01 position and settle at the 02 position. All scram times for this rod were within prescribed limits but one scram at 1000 psig showed 90% insertion to be 3.6 seconds.
(Technical specifications 3.2.B.3. require the average of the scram insertion times of all operable. control rods to be no greater than 5.00 seconds for 90%
insertion).
Control rods 14-35 and 42-27, which did not latch on scram 42, did latch at 00 on this scram..
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- Inquiry Memorandum 219/70-D.
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- Inquiry Memorandum 219/70-E.
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Date cause April 19, 1970 Automatic scram from 7% power from reactor neutron (No. 44) monitoring. The reactor was being taken off the line for control rod work and the operator was inserting rods with the IRM detectors in range six., As the detectors started to move in, the flux as seen by the j
IRM's increased faster than the operator could change 1
the range switch position. To avoid this condition
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in the' future, the IRM detectors will be inserted very slowly with the IRM's in range nine (9) after the APRM's reach approximately 10%. The mode switch should be placed in "startup" mode before the APRM's reach their downscale trip point'as called for in the shut-down procedure.
E.
Primary System
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1.
Primary System Chemistry
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Mr. McCluskey informed the inspectors that the main condenser had been opened and inspected.during the April - May, 1970 rod work outage. Only minor, insignificant amounts of crud was observed and removed. During the last two entries into the main condenrer, significantly more crud was observed and removed.
Based on the above information it appears that the crud buildup and accumulation in the system has substantially decreased.
2.
Reactor Vessel Nozzles During the March 18 - 20, 1970 inspection, Mr. Caphton reviewed the Nine Mile Point Nucicar Station's core spray nozzle cracking problem and urged JC to review the Nine Mile Point experience as it may be applicable to the OC-1 plant. At that time Mr. McCluskey informed Mr. Caphton that Mr. Rees, Project Manager, OC-1, was closely following events at Nine Mile Point and that appropriate action would be taken by JC.
This appropriate
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action was stated to include a nozzle stress evaluation and the adequacy of seismic restraints.
Subsequently, Mr. I. Finfrock. informed the inspector that the above checks had been made and that during the rod work outage (April 21 - May 28, 1970) the exterior cladding on five reactor nozzle safe ends was PT inspected with the following results:
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South Core Spray Nozzle - The initial PT inspection disclosed several indications of questionable relevancy due to the as-velded (rough) surface condition of the overlay deposit. The overlay surface was polished by light grinding and PT checked again and all indications had vanished.*
b.
North Core Spray Nozzle ** - Based on the experience with the south core spray nozzle safe end, the overlay cladding on this safe e'nd
- Acceptance criteria in accordance with the requirements of ASME,Section III.
- Inquiry Memorandum 219/70-F - LINEAR INDICATIONS IN WELD OVERLAY ON NORTH CORE SPRAY SAFE END.
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t was lightly polished prior to PT inspection. The FI check disclosed several linear defects in the overlay cladding covering a non-sensitized portion of the safe end.
(See Addendum 3 attached). A l
boat sample was taken in the area of one of the more significant linear defects. The metallurgical analysis of the boat sample together with the resolution of this issue, is discussed below.
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Control Rod Drive Hydraulic Return Nozzle - The overlay cladding on i
the safe end was PT checked without prior polishing and no relevant g
indications were noted.*
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Isolation Condenser Nozzle Safe Ends (Steam Line Nozzles) - Both isolation condenser nozzle safe ends were PT inspected without prior polishing and no relevant indications were noted.*
Metallurgical Analysis of Defects on North Core Spray Nozzle Safe End Cladding and Resolution of Issue As discussed above, a PT check of the overlay cladding on the subject safe end disclosed several linear indications up to'3/8" in length.
As a result of these findings, Mr. W. Reinmuth, Senior Reactor Inspection Specialist, CO:HQ and Mr. A. Holt, Chief, Materials & Metallurgical Branch made a special inspection of the OC-1 site on April 30, 1970. The results s
of their inspection have been previously reported.** A boat sample which included a linear defect in the 12:00 o' clock position was removed and metallurgically examined by GE-APED. GE concluded that the cladding was Inconel and that the cracks (microfissuring) were caused by weld solidification during the application of the cladding. The methods and results of the GE investigation were reviewed by Messrs. Schmidt, MPR Associates, and Wiley, Southwest Research Institute, who were acting in the cap'acity of consultants to JC.
They both concurred with the GE findings. GE and JC representatives discussed their investigations and conclusions at a meeting with DRL in Bethesda, Maryland on May 14, 1970.
Contour grinding of the exterior cladding on the north core spray nozzle '
safe end was accomplished to remove all linear defect = and to blend the area disturbed by the removal of the boat sample. Tn was reported to involve the removal of metal from the area in questi4d cu 3600 of the pipe surface and in some cases this may have removed aA1 of the Inconel overlay cladding and may have penetrated somewhat into the non-sensitized base material. Due to the relatively inaccessible location of this nozzle and the somewhat irregular contour grinding, GORB questioned the adequacy of the remaining wall thickness and directed that a determination of the wall thickness be made. The GPU quality assurance group (Messrs.
B. Avers and N. Goodenough) developed and implemented a procedure to determine the wa11' thickness by radiography. The measured wall thickness by this method was stated by the licensee to exceed code requirements.
- Acceptance criteria in accordance with the requirements of ASME,Section III.
- Memorandum (with enclosed report), Engelken to Morris, OYSTER CREEK CORE SPRAY SAFE END OBSERVATIONS, dated May 21, 1970
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The adequacy of this measurement technique was questioned by CO:I and Messrs. Tillou and McDermott made a special inspection to the GPU - JC hame office located in Parsippany, N. J.,
to review this matter on May 22, 1970. Discussions were held with Mr. B. Avers, QA Manager, CPU, Mr. N.
Goodenough, QA Engineer,- GPU, Mr. W. Hirst, Chairman, GORB, GPU, Mr. I.
g Finfrock, Manager of Nuclear Generating Stations, JC, Mr. J. Moore, Piping Engineer, GPU and Mr. D. Hetrick, Operations Supervisor, OC-1.
The discussions focused on (a) determining that an adequate wall thickness i
did remain, (b) reviewing the radiographic procedure that was utilized,
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(3) reviewing the procedures qualification, and (d) determining the 3
adequacy of JC-GPU's interpretation of the results.
The procedure used was titled " Procedures for Determining Pipe Wall Thick-ness by Radiography" which was dated May 15, 1970 and approved by Mr.
Avers. The procedure had been qualified by the same technician who took the field shots on the north core spray nozzle safe end. Good agreement was noted between the two pipe wall sections radiographed in the qualification shot and the micrometer measurements of the same sections.
The largest error noted was approximately.004 inches on a wall thickness of approximately.450 inches.
Mr. Finfrock and Mr. Moore informed the inspectors that GE and GPU had independently calculated and arrived at similar findings of the required wall thickness for the north core spray nozzle safe end. This was stated to be 0.401 inches.* The minimum measured value of the four sections of the pipe wall which were radiographed were stated by Mr. Goodenough to be 0.419 inches by his interpretation of the film.
CO (Tillou) made measurements in a similar manner from the radiegraph and determined that the minimum wall thickness shot was approximately 0.410 inches.
Mr. Goodenough was asked why the more widely. accepted method of UT was not used to determine the wall thickness and he stated that the compound contour surface condition produced by the generally tangential grinding did not permit a UT check. He was also asked how he assured himself that the thinest sections of the pipe wall had been radiographed. He informed the inspectors that a visual inspection of the safe end was made and that the deepest appearing areas were selected to be radiographed.
The licensee has reported ** both the results of PT inspection of the reactor vessel nozzle safe ends and the resolution of the linear defects observed in the north core spray nozzle, i
Based on the observations made during the inspection, that is, a small i
measured excess in pipe wall thickness over code requirements, the method of determination of pipe wall thickness (radiography) and its associated errors (expected accuracy was stated by Mr. Goodenough to be j
plus or minus 10 mils max), and the method of selection of wall sections j
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- Basis for this determination was ASA B31.1-1955, Paragraph 122
- Letter frvm Mr. I. Finfrock, Jr., Manager of Nuclear Generating Stations, JC to Dr. P. A. Morris, Director, DRL dated June 16, 1970.
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to be shot (eye-ball check for the deepest appearing depressions), the assigned inspector contacted the station superintendent subsequent to the inspection'and stated that it appeared prudent to perform a more t
precise and thorough wall thickness determination.
Mr. McCluskey stated
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that he would consider this item. This matter will be pursued further i
during the next routina inspection.
i 3.
Leaking Primary System I tstrument Line l
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Main Steam Isolation valve Leakage 1
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(See Section K.)
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5.
Isolation Condenser Instrumentation Changes The inspectors discussed with Messrs. Riggle and Hetrick an occasion when the isolation condensers had inadvertently isolated themselves from the system following the first initiation of the system under load conditions.
This occurred during startup testing in the fall of 1969.* The inadvertent isolation was reported to be caused by a transient.high flow spike as seen by the condensate flow instrumentation located in the condensate return line to the reactor. This instrumentation is utilized to monitor for.
line breaks in the isolation condenser and associated pipes and to ' initiate a valve closure action (two steam valves and two condensate valves) for each isolation condenser. To prevent recurrent isolations, a five-second time delay was added to the circuitry for both the steam flow sensing and the condensate flow sensing instrumentation. This time delay functions to allow for a higher flow than the trip settings of the instruments to persist for up to five seconds before a valve closure signal is initiated.
Subsequent to the inspection the licensee reported during a telecon with CO:I on July 1,1970**, that following the inadvertent isolation discussed above, the trip point had been raised from 27" 4 P H O to 59" aP H O 2
2 as per a GE instruction letter. During the telecon, Mr. McCluskey was informed that technical specifications require a setting of -E 27" 4 P H O2 and therefore he was considered by the inspector to be operating the plant in noncompliance.
Mr. McCluskey stated that GE had recently provided JC with calibration data for this instrumentation and that it was JC's conclusion that the technical specifications were in error.
He-was advised to immediately contact DRL to resolve this question. The licensee subsequently reported, on July 2,1970, that the setpoint on-line instrumentation was returned to the required 27" 4 P H 0.***
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- 0C-1 Semi; annual Operations Report for May 3, 1969 - December 31, 1969, dated Apr,il 2, 1970.
- Inquiry Memorandum 219/70-H " Isolation Condenser Instrumentation Setting -
Noncompliance.
- TWX to USAEC, HQ, Germantown from I. Finfrock, JC, dated July 2,1970.
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This issue will be reviewed during the next inspection and the appropriate enforcement action will be taken at that time.
F.
Reactivity Control and Core Physics b
1.
Excess Flow Check Valve Closure During Operation. Loss of Protective Instrumentation - Item of Noncompliance 1
The licensee previously reported * (Scram No. 31 dated December 31, 1969)
I that the reactor scrammed from a low water level trip from a malfunction of the 3-element feedwater controller caused by a stuck excess flow check valve on a steam flow sensor. A more thorough discussion of the scram is documented in C0 Report No. 219/70-1 (Scram No. 8).
The most recent review of this scram included discussions at the site and a review of records and logs. The review disclosed that the cause of Scram No. 31, i.e., stuck closed excess flow check valve, may not have been established until after resuming operations on January 1, 1970.
(See CO Report No.
219/70-1, Scram No. 8, which is a summary of the JC scram report).
In discussions with Messrs. Hetrick and Riggle on this subject, there was a s trong indication that the cause for the excess flow check valve going closed was the opening of an instrument bypass valve during periodic testing on one of the instruments connected to the sensing line in question, but this could not be definitely. established. Closure of the excess flow check valve in the subject sensing line or leakage through a bypass line could result in inaccurate sensed 4 P's and change the trip
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setpoints (in effect) for the following instruments:
a.
Four 6 P struments to initiate main steam isolation valve (MSIV) closure on receipt of a high dP across the venturi installed in the main steam line. These four instruments (and the instrument in
- b. below) share a common single pair of sensing lines and provide major steam break protection. Normal 4 P signal (100% steam flow) is in the range of 47 - 51 psi A P versus the setpoint for MSIV closure 63-1/2 pounds (120% of rated steam flow). A bypass valve being opened would decrease the measured a P, and in effect elevate the trip setpoint in an excess of that analyzed for in a major steam line break.
b.
One o P instrument to provide steam flow input to a steam flow totalizer, which in turn provides the steam flow signal to the 3-element feedwater controller. A bypass valve open on one of the instruments would result in a lower measured A.P and transmitted steam flow signal. This would result in the 3-element feedwater controller demanding less than the required feedwater flow and a low reactor water level scram could be expected to result. A sudden closure of an excess flow check valve caused by flow through a bypass valve could also affect the feedwater controller in a similar fashion, only a more sudden reaction could be expected.
- 0C-1 Semi-annual Operations Report for May 3, 1969 - December 31, 1969, dated April 2, 1970.
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' The shif t supervisor's log was reviewed by the inspector and noted to l
contain the following information:
Entry at start of 0000-0800 shift January 1, 1970 recorded the reactor was critical.
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Entry at 0204 recorded reactor power at 257., that the main steam flow indication had gone to zero and that it was necessary to tap
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on the excess flow valves bef re measured steam flow started to l
increase on the control room recorder.
Note: (By Writer) This last entry could be explained by the measured l
/J P across the steam line venturi increasing (as power level and steam flow increased) to the point where sufficient flow was developed i
through an open instrument bypass valve to close the excess flow check valve.
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The control room operator's log was also reviewed by the inspector and noted to contain the'following:
Entry at 0254 on January 1, 1970 recorded that reactor steam flow indication was not following and it appeared that the excess flow check valves may be shut._ Attempts were made to reset the excess flow check valves by tapping on valves and following this the steam flow signal began increasing on the recorder.
I Based on the above it appears that Scram No. 31 resulted from an instrument bypass valve being opened and subsequent closure of an excess flow check valve. Also. based on the above information and discussions j
with Messrs. Riggle and Hetrick, it is the inspector's conclusion that i
JC may have conducted an inconclusive review of the cause of Scram No. 31 (December 31, 1969) before resuming operations.
Furthermore, during the period of operation from 0000 to 0254 on January 1,1970, a limiting l
condition for operation was not met (Reference Table 3.1.1 of technical specifications, items 2 and 3) in that the major steam line break 4
k protective instrumentation trip points were (in effect) elevated to a higher setpoint than specified. This was caused by the reduction in the measured steam flow A P or a zero measured A P due to either an instrument i
bypass valve being open or the excess flow check valve being closed respectively.
1 The inspector's observations were discussed during the exit interview with Mr. McCluskey on May 21, 1970. The inspector commented ~ that op' era.
tion of the reactor on January 1,1970 without the instrumentation speci.
ficd in a. above was in violation of a limiting condition for operation (Technical Specification, Table 3.1.1, item B.2 or 3) and therefore was an item of noncompliance.
Mr. McCluskey did not comment.
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Leaking Primary System Instrument Line. Loss of Protective Instrumentation -
Item of Noncompliance Region I was contacted by telecon on March 25, 1970 by Mr. T. McCluskey, Station Superintendent, OC-1, who reported that a limiting condition for operation had been violated.* The event involved a Swage-lok tube fittit4 leak on a reactor high pressure scram switch sensing line. During l
attempts to tighten the fitting the line became disconnected. The leakage i
through the open line exceeded the upper limit of the excess-flow check valve located in the leaking line (designed to protect against;such down-stream leaks).and the excess flow check valve closed.
Mr. McCluskey stated at the time that no scram protection was lost but the event did result in*
,k a loss of protective instrument redundancy.
Mr. McCluskey also stated that preparations had been made to shut the reactor down but that the fitting was repaired and the instrument returned to. service before the shutdown was initiated.
This was reported to require approximately 50 minutes. JC has submitted a report of this event to DRL** in accordance with Section 3.C.1 of the license.
A meeting was held at CO:HQ on April 30, 1970 involving Messrs. D. S.
Sullivan, DRS, R. McDermott, C0:1, S. Bryan, CO:S&PB, and V. Thomas, CO:TSB, to discuss the application of excess flow check valves in instrument lines and to review the JC event discussed above. The results of this meeting were discussed in Mr. H. Denton's memorandum to Mr. J. P. O'Reilly, 4
dated May 5, 1970.***
As there was insufficient information (schematic drawings) to perform a complete review at the April 30, 1970 meeting
. discussed above, Mr. Sullivan visited the OC-1 site on May 19, 1970, during the scheduled inspection, to assist the inspector in completing the review of the March 25, 1970 instrument line failure.
The site review consisted of an inspection of as-built schematic drawings and discussions with Messrs. D. Hetrick and W. Riggle. The facts 1
disclosed during this review were as follows:
a.
The problem occurred during periodic testing on REO 3 C (High Reactor Pressure Scram Switch) when a leak developed in a Swage-lok fitting upstream of the instrument root valve. An attempt to tighten this fitting resulted in the instrument line becoming disconnected. This resulted in the closure of the excess flow check valve which then removed the reactor pressure from the sensing instruments on the instrument rack.
(Note:
Several instrument sensors utilize a common sensing line in the 0C-1 plant. During the period the excess check valve was closed and the fitting was leaking, a zero pressure signal was seen by the affected sensors.)
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b.
The sensors involved in this event are outlined in two memoranda.****
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\\*InquiryMemorandum219/70-C,OPENINGOFREACTORHIGHPRESSUREINSTRU LOSS OF SAFETY SYSTEM REDUNDANCY.
- Letter from Mr. George Kelcec, Manager of Generating Stations, JC to Dr. P. A.
kMorris,DirectorofDRLdatedApril3,1970.
- emo from Mr. H. R. Denton, Chief, TSB, CO:HQ to Mr. J. P. O'Reilly, Chief, t
RI&EB, 00:HQ, entitled INSTRUMENT LINE - FLOW CHECK VALVE APPLICATION, dated May 5, 1970.
- Inquiry Memorandum 219/70-C dated March 27, 1970 and memo from H. R. Denton, CO:HQ to J. P. O'Reilly, CO:HQ, entitled INSTRUMENT LINE - FLOW CHECK VALVE APPLICATION, dated May 5, 1970.
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The excess flow check valves in use at OC-1 were reported by Mr. Riggle to be of-the ' eccentric seat design to provide for a controlled 2 gpm leakage at 1000 psi 6 P.
Mr. Riggle reported that in the March 25, 1970. break, the_ excess flow check valve' did not exhibit controlled i
leakage and that it stopped the'1eak almost entirely. Eccentricity L
of the seat was not obvious to the inspectors when an inspcstion was made of a spare excess flow check valve. Valve body markings on f
installed excess flow check valves only indicate the direction of flow to initiate closure, and specify 4 gpm maximum (implying this is-i i
the flow that would result in overcoming the. spring force in shutting
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the valve).
d.
Excess flow check valve closure in this event resulted in the loss in the ability of two of the four autodepressurization valves to auto-matica11y open on high reactor pressure. Manual (remote-electric) control of these autodepressurization valves was retained as was the i
ability of these valves to automatically provide the blowdown function.
f e.
Loss of redundant instrumentation also resulted for (1) high pressure-scram, (2) high pressure isolation condenser actuation, (3) condenser low vacuum and turbine trip scram bypass, (see f. below), (4) core spray valve permissive, and ~(5) triple low level autodepressurization.
l One of the two reactor level sensors used in the 3-element (level -
steam flow - condensate flow) controller of the feedwater control system was also lost as was the sensor for the reactor pressure indicator - transmitter. The loss of the level transmitter for.the 3-element feedwater controller did not affect plant operations as.the-i alternate transmitter had been selected for control at that time.
If this transmitter had been selected, the control system would have -
called for a large condensate flow as the level transmitter would have seen a zero reactor level signal.
l f.
Fortuitously, no scram protection was lost. as the reactor mode switch was in the "run" mode which is the normal condition for operating at reactor power in excess of 107..
Total-scram protection would have been lost had the mode switch been in the "startup" or " refuel" mode, for both the low condenser vacuum and the main steam isolation valve closure scram features. The zero pressure signal (anything less than 600 psig is sufficient) will actuate relay contacts to bypass these scram features in total when a pressure of less than 600 psig is sensed and the reactor mode switch is in the. "startup" or " refuel" posi-tion. When the mode switch is in the "run" position, the scram' j
features are not bypassed for any reactor pressures.
g.
The ability to automatically actuate both Isolation Condensers on high reactor pressure was negated due to a design error in the relay 1
matrix. The as-built relay matrix logic was confirmed to be I
designed incorrectly based on a review of the schematic drawings.
JC and GE were currently re. viewing the required design change to correct.the situation. The required change was reported to involve an' interchange of two wires.
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4 Mr. Riggle and Mr. Hetrick were asked why JC.had not reported (during the telephone conversation on March 25, 1970) the loss of automatic over-pressure protection for two to four autodepressurization valves.
Mr.
Hetrick stated cl:at the lack of reporting of this facet had been an over-
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sight on JC's part.
Based on discussions in the technical specifications' basis, page 2.3-4, paragraph 2 and information presente,d in the FSAR Section IV, " Reactor Primary System" 2.1, the inspector's evaluation is that the event of the k
failed fitting resulted in noncompliance with the intent of the specified Limiting Safety System Setting. Although the sensing instrument setpoint was, or may have been correct, i.e., le.as than or equal to 1125 psig, no valid high reactor pressure votild have actuated two of the four relief (autodepressurization) valves. This item was discussed in the exit inter-view and Mr. McCluskey was informed that the inspector considered the loss of two (of four) r~elief (autodepressurization) valves to relieve on a high reactor pressure and the failure to report same as items of noncompliance with the technical specifications.
Mr. McCluskey acknowledged these as items of noncompliance.
3.
Control Rod Drive Problems - Noncompliance Item One hundred and thirty-seven (137) of the 138 installed control drives were rebuilt as were three spare drive units during the reactor shutdown for that purpose from April 19, 1970 to May 22, 1970.
Mr. D. Pomeroy, TSB, CO:hQ performed an assist inspection and the results of this inspection are documented in his June 5, 1970 report.
(See Addendum 4).
One item of noncompliance was identified in Mr. Pomeroy's review. The rod drives were not exercised on April 15, 1970, as required s'. ten there are two inoperative rods (Technical Specification 4.2.D).
This item of noncompliance was identified during the exit interview with Mr. McCluskey on May 21, 1970.
Mr. McCluskey acknowledged that this was a technical specification violation and stated it probably resulted from the busy operating schedule in attempting to restart the reactor following a scram on April 14, 1970. Following this scram the inoperability of the second j
control rod was established which then required daily exercising per the technical specifications.
4.
Scram Reset Circuitry Problem l
A problem was previously reported (Nine Mile Point Nuclear Station)* when 10 (of 129 total) control rods had missed being scrammed to the full in i
position by one notch and one control rod had missed by two notches. As a result of the NMPNS investigation of this problem, two modifications were made to the scram circuitry at that facility to prevent the auto reset of momentary transient scram trips and also to eliminate the 1
potential for an operator to reset a scram condition before the control
- Letter to.Dr. P. A. Morris, Director, DRL from Mr. A. Burt, Station Superintendent, NMPNS, Nitgara Mohawk Power Corporation, dated November 18, 1969; and CO Report No. 220/69-16, Section F.2.
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rods were fully inserted. Based on this information,'JC management was appraised of these developments and they.in turn contacted GE to request a proposed design change for the OC-1 facility.
The subject of timeliness in implementing this change was discussed during the exit interview at the time of the March 18-20, 1970 inspection.*
Mr. McCluskey stated during that exit interview, that the originally 1
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proposed design change provided by CE was returned to GE for clarification L(
as the method for performing.the change was not clear to the OC-1 i
personnel. Mr..McCluskey also stated that this matter would be given priority attention and that PORC would review the change when GE provided clarification for the proposed changes. This item was also discussed at the Regulatory-JC meeting held May 14, 1970 to discuss results of JC's investigation of control rod drive problems.
During this inspection it was determined that the change proposed by GE for the OC-1 facility is similar to the modifications made at Nine Mile i
\\ Point Nuclear Station.
Mr. McCluskey stated that the GE proposal is i
scheduled to receive review by PORC, prior to the changes being made. JC I
at this time still has reservations on installing the fast action relays to prevent the auto scram reset for short (less.than approximately 10 milliseconds) duration trip signals. This issue was reported to still be under review by JC and a decision on this matter would be made in the near future. Mr. Riggle stated that he had conducted relay coordination studies of the scram relays and had reservations pertaining to the installation of the fast acting relays.
Mr. McClusk*ey also stated that-if the change is not accomplished within three months, the 00-1 staff will have.to so justify to the GORB.
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5.
Instrument Line snubbers i
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Mr. Riggle informed the inspectors that snubbers (instrument pulse dampin devices) had been installed in the following locations:
a.
Steam line measuring racks for both main steam lines - This involves fou)OPinstrumentspersteamlineforthemajorsteambreak protection feature and a A P transmitter used in the 3-clement feed-water controller.
b.
Isolation condenser. condensate line break d P instruments.
The snubbers are not readily observable when installed in the sensing lines as they appear in normal fittings approximately 3/4" to 1" long. The snubbers installed at 0C-1 in the above-mentioned applications are manufactured by Chem-Quip. Mr. Riggle informed the inspectors that the type installed at OC-1 have an E perosity density (scintered metal) and are designed to pass 3 cubic feet per hour (cfh). at 1 psi air A P.
They are specified by the manufacturer for use with water and light oil'.
Mr. Riggle also ' informed the inspector.that the snubbers hrd been installed to eliminate instrument " bounce".
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- C0 Report No. 219/70-3, Management Interview 2
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The inspector discussed the use of snubbers
.h Mr. McCluskey both during the exit interview and subsequently elecon. The inspector's position was that there is a real potential sr crud to plug the pores of the scintered metal and result in increased instrument response
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times.
Mr. McCluskey was asked if PORC had reviewed and approved the use of snubbers and he replied "yes".
The inspector asked Mr. McCluskey-if PORC had considered (a) the potential for increasing the response g
times dut *o plugging, (b) performing measurements of response times 1;
af ter inttial installation and periodically thereafter, and (c) periodically
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backflushing instrument lines to minimize the potential for plugging.
Mr. McClusicy stated he was uncertain what was reviewed but that he L
would look into the matter. The inspector stated that if these considera-tions had been incorporated in the review, he saw no evidence that items (b) and (c) above were being implemented. The inspected requested Mr. McCluskey to advise the GORB of the stated concerns for the purpose of a GORB review of this matter.
Mr. McCluskey indicated that GORB will be so informed. This item will be reviewed during the next routine inspection.
H.
Power Conversion System Two of the four scrams occurring during the inspection period were initiated by or associated with turbine control steam pressure regulator malfunctions.
In the last instance, during a transfer from the mechanical to the electrical pressure regulator, system instability occurred and resulted'in system pressure swings which.
eventually _ initiated a system scram from low reactor water level.
It was reported that the preasure increased to the point where an electromatic relief valve functioned. JC has contacted GE on this problem and it is reported that-the pressure control system,is under current review by both JC and GE.
The details of both of these scrams are discussed in Section C (Scrams No. 42 and No. 43).
Followup review of the GE findings and recommendations will be made during the next routine inspection.
K.
Containment 1.
Main Steam Isolation Valve (MSIV) Performance - Item of Noncompliance Valve leakage rate tests were conducted during the April - May 1970 rod work outage and two (of four) failsd to meet the specified leakage rate. The results of these tests and the corrective actions taken by JC have been reported by the licensee.* Testing'was accomplished in accordance with a prior license commitment.**
- Letter to Dr. P. A. Morris, Director, DRL from Mr. I. Finfrock, Jr., Manager, Nuclear Generating Stations, JC dated June 3,1970
- Letter to Dr. P. A. Morris, Director, DRL from Mr. George Kelcec, Manager of Generating Stations, dated March 20, 1970.
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\\' s Discussions were held with Mr. D. Ross, Technical Supervisor, OC-1 to obtain the details of the problems experienced with the valves.
He provided the following information:
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a.
The valves were rested on April 21, 1970 after only a single closure following the reactor shutdown on April 15, 1970 The north set was f ound to be tight when a 20 psig air pressure was 3
applied betweec the valves. By the same method the south side g
(one or both) were determined to be leaking in excess of technical y
specifications.'
f b.
On April 27, 1970, work began on the south inboard valve (NSO3B) after a 16 psig pressure applied to the reactor vessel disclosed that the valve leaked in excess of technical specification limit. No i
measured leakage rate for this test was provided by Mr. Ross.
Only 16 psig could be built up in the reactor vessel due to leak-through on an electromatic relief valve on the south steam line.
c.
On May 13, 1970 the air and hydraulic cylinders on NS03B were dis-assembled and inspected. A mixture of rust, dirt, and moisture was removed from the bottom of the air cylinder.
d.
On May 6, 1970, Mr. J. Festa, Atwood Morrill representative (main MSIV supplitr) was at the site to review the problem.
He recommended increasing the clearances on the pilot stem in the vicinity of the j
guide tube as a corrective measure. The pilot stem was removed and measured to have a 30 mil total indicated run-out.
Following repairs on NS03B, both inboard valves were tested satisfactorily on May 8, 1970.
e.
The south outboard MSIV (NSO4B) was tested and found to leak in excess of technical specification requirements. This valve was dis-assembled and found to have a total indicated run-out (bowing) of 60 mils which was corrected to 7 mils. Ten mils was taken off the pilot stem in the vicinity of the guide tube, and 5 mils removed from the poppet cap.
f.
The final valve tests were completed successfully and the measured leak rates were reported as follows:
(11.5 SCFE is the T/S ILmit)
NS03A 4 0.1 SCFH NS03B 4 0.1 SCFH NSO4A < 0.3 SCFH l
NSO4B = 1.5 SCFH To date, the history of performance of the OC-1 MSIV's has been poor, as outlined below:
a.
Prior to initial criticality in' June 1969, 00-1 had considerable difficulty in meeting technical specification requirements for leakage limits.*
- CO Report No. 219/69-5, Section K.1.
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During startup testing on December 9, 1969, steam pressure buildup was noted downstream of the MSIV's during periods when.the MSIV's were closed. This pressure buildup was in a section' of the steam '
3 line.between the MSIV's. and the turbine stop. valves.*. Subsequently 1'
the plant was shut down on February 1, 1970,-and the leakage for three of the four valves were determined to be well in excess of l
bg technical specification requirements.
CO made a special inspection **
to review the results of this testing.
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c.
The most recent failure discussed above when two (of four) valves J
1eaked in excess of limits.
In the most recent instance, when the MSIV's failed to meet leakage specifications, the licensee did not report this item to DRL within the 10-day period required by 3.c. (1) of the license. The determination was made on April'21, 1970 and not reported on until June 3, 1970 This is considered to be an item of noncompliance. This C0 position was communicated to Mr. McCluskey, Station Superintendent, during the exit interview conducted.on May 21, 1970 Mr. Ross and Mr. McCluskey stated that they had considered that they would issue the report following the results of the investigations as they had considered that this would have been of more benefit. - They also stated the reactor had remained down following the findings of an excessive. leakage. The inspectors outlined for Mr. McCluskey the rationaleron which reporting requirements are based and the possible safety application to other plants.
At present, JC only intends to recheck the MSIV's in accordance with the prior commitment,*** that is, to test at every opportunity when the plant is shut down and cooled down until the next refueling outage.
U.
Miscellineous 1.
_Pumperete The problems experienced with concrete transported through aluminum pipe at other facilities was discussed with Mr. McCluskey during the March 18-20, 1970 inspection.**** Mr.'McCluskey was appraised of the findings of up to 507. reduction in the compressive strength of concrete transported in this manner.
It was requested by the inspectors at that time that JC review the OC-1 construction records to determine if this method had been used.
- Letter to Dr. P. A. Morris, Director, DRL, from Mr. G. H. Ritter, Vice President, JC, dated December 24, 1969.
- C0 Report No. 219/70-2.
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- Letter to Dr. P. A. Morris, Director, DRL from Mr. George Kelcec', Manager L
of Generating Station, JC dated March 20, 1970.
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-****C0 Report No. 219/70-3, Management Interview Item 1.
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' - During this inspection, Mr. McCluskey informed the inspectors that this method of concrete transporting had been used at OC-1.
A letter (Rees to McCluskey, dated April 7, 1970) pertaining to this subject,.was reviewed s
j by the inspectors. The letter stated that Burns and Roe (B&R) had reviewed the 00-1 construction records and determined that aluminum pipe
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wasysedtotransportconcretetoanareaunderthedrywellinsideofthe drywell support skirt. After this concrete was poured and set, the steel support skirt was cut loose from the drywell. The letter further stated,
y that the concrete in question had a design strength of 4000 psi and was completely surrounded by other concrete and it would be impractical to take e
core sampics at this time. The concrete surrounding this area is probably in excpss of five feet thick and would be in addition to the steel skirt which was left in place. To the best of B&R's knowledge, this was the only concrete placed with aluminum pipe.
Mr. John Archer, Lead Engineer, B&R, did not feel that aluminum pipe transported concrete presented a problem in this situation, as the concrete in question is completely contained by other high strength concrete.
In addition, the concrete in question is in compression and since it is contained on all sides, there does not seem to be any failure mode possible.
V.
Reliability Information 1.
Failure of two (of four) of the main steam isolation valves to meet leakage limits was detected during the April - May, 1970 rod work outage.
(See Section K.)
'2.
A leak developed in a Swage-lok fitting on a primary system instrument sensing line during plant operation and resulted in removing some plant protective features.
(See Section F.1.)
i 3.
The closure of an excess flow check valve in a primary system instrument sensing line during operation resulted in removing a portion of major J
steam line break protective instrumentation from service.
(See Section F.2.)
4.
Extensive problems experienced with control rod drive assemblies are discussed.
(See Section F.3,)
)
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