ML18142B666

R. E. Ginna, Unit 1 - 07/08/1969 Memo Inspection Report 05000244/1969009

ML18142B666
Person / Time
Site:	Ginna
Issue date:	07/08/1969
From:	O'Reilly J US Atomic Energy Commission (AEC)
To:	Boyd R US Atomic Energy Commission (AEC)
References
IR 1969009
Download: ML18142B666 (33)

Text

glCI ro a UNITED STATES I O e ATOMIC ENERGY COMMISSION O WASHINGTON. D.C. 20545 o

~ALIIS 0> JUL 8 $ 68 R. S. Boyd, Ass is tant Director for Peactor Projects Division of Reactor Licensing (2)

ROCHESTER GAS AND ELECTRIC CORPORATION, R, E. GINNA, CO REPORT NO. 244/69-9 The enclosed report of an inspection visit to the subject facility on May 19-21, 1969, is forwarded for information and for possible action.

The report provides iniormation concerning the satisfactoxy resolution of a number of problems that wexe detected during our intensive quality assurance inspection, CO Report No. 244/69-1. Ile are unable to obtain conclusive verification that improper welding rods were not used in the field fabrication of important piping. Oux consultant, Dr. R. Gilliland, Parameter, Inc., believes, and we concur, that a program of nondestructive inspections should be performed periodically to monitor the integrity of the primary piping systems. This program should be oriented to detect fatigue cracks that could result from periodic thermal cycles if improper dissimilar filler metal was used for welding. Accordingly, we recommend that DRL consider incorporation of such a program in the Technical Speci-fications. In our view, an acceptable program would require both volumetric and surface inspections.

The status of the other problems that were highlighted during the quality assurance inspection is as follows:

a. Contouring of in-core instrument penetrations - satisfactory con-touring has been accomplished as described in CO Report No. 244/69-8.

b. Battery room doox - the licensee stated that the PSAR will be changed to reflect the installed one and one-half hour door. [?e will consider this as an outstanding item until the licensee effects thc change.

c~ Information concerning the satisfactory resolution of arc stxikes and weld spatter, hydrogen, generation and battery room ventilation will be obtained and provided in subsequent inspection reports.

Original signed hg J. G. Keppler J. P. O'Reilly, -Chief Reactor Inspection and Enforcement Branch Division of Compliance

Enclosure:

CO Report No. 244/69-9

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R. S. Boyd

, cc: '/encl, E. G. Case, DRS S. Levine, DRL (6)

D. J. Skovholt, DRL (3)

L. Kornblith, CO J. G. Davis', CO:II

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B. H. Grier, CO:IIX D. X. >lalker, CO:IV R. 0J. Smith, CO:V N. C. Moseley, CO: I, w/o encl

~Keg Central Pile

U. S. ATOMXC ENERGY COMMXSSXON

, REGXON X DIVISION OF COMPLIANCE

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4 Report of Inspection CO Report No. 244/69-9 Licensee: Rochester Gas and Electric Corp.

License No. CPPR-19 Category B Date of Inspection: May 19-21, 1969 Date of Previous Inspection: April 29-30, 1969 Inspected By: c D. M. Hunnicutt, Reac r Inspector Reviewed By:

N. C. Moseley, Senio Reactor Inspector Proprietary Items: None SCOPE An announced inspection was made to the Robert E. ,Ginna Nuclear Power Plant Unit, No. 1 construction site on May 19-21, 1969. The Ginna Nuclear Power Plant is a 1300 Mwt unit under construction near Rochester, New York.

SUMMARY

Eighteen items identified during'the special quality, control in-spection and reported in CO Report No. 244/69-1 have been resolved.

The plant has experienced two electrical outages. .Both occurred while the diesel generator units were out of service for investiga-tion following automatic connection to the distribution grid when an incorrectly installed relay indicated the bus was "dead" in-stead of indicating the bus was energized. The Reactor Coolant System was at normal operating temperature when the first elec-trical outage occurred. This outage lasted for approximately forty minutes.

e' Reactor Coolant Pump No. 1-A indicated erratic cooling water flow to the labyrinth seals. The problem is to be investigated and /he seals changed out. The graphitar bearings will also be inspected.

The main coolant relief valves received adequate nondestructive testing and the acceptance testing was in accordance with the re-quirements stated in BGPV Code,Section III.

A comprehensive inspection by the Applicant indicates that no valves in safety oriented systems are carbon steel or contain carbon steel internals.

The shafts on two of the high pressure safety injection pumps could not be turned by hand after the pumps had been operated.

The pumps were locked out and the Applicant requested technical assistance from the pump manufacturer.

The Code stamps have been attached to the steam generators in compliance with Section III of the BSPV Code.

The 7 nut plates on the channel heads of the steam generators were radiographed, ultrasonically tested and heat treated.

The containment vessel leak rate test results were:

Leak Rate N 60 psig 0.0238% + 0.0168% per day Leak Rate N 35 psig 0.0053% + 0.0180% per day Several vanes in the air circulating system broke off during the containment vessel overpressurization and leak rate testing program.

Followup action is in progress to determine the cause(s) for this equipment failuge.

Inspection by RGGE personnel indicates that gaskets have been in-stalled in all instrumentation housings to preclude moisture from affecting the operation of instruments in the containment vessel.

The Applichnt,.has> Completed;plans,"o install a-.bypass:",line'round.."

the, instrument~air'..Supply "

DETAXLS A. Persons Contacted Mr. John Arthur, Project Manager, RG&E Mr. Charles Sundstrom, Project Engineer, RG&E Mr. Charles Mambretti, Mechanical Engineer, RG&E Mr. R. Latz, Electrical Inspector, RG&E Mr. Charles Platt, Plant Superintendent, RG&E Mr. Earl Ford, Ginna Project Manager, Bechtel Corp.

Mr. John Gilbert, Field Service Engineer, Bechtel Corp.

Mr. L. Elliot, Assistant Resident Engineer, Westinghouse Mr. E. U. Powell, Ginna Project Manager, Westinghouse Mr. James Dolan, Le'ad Engineer, Westinghouse Mr. J. R. Zahors'ky, Senior Project Engineer, Crosby Valve Co.

Mr. W. C. Sommers, Nuclear Engineer, Gilbert Associates, Inc.

B. Resolution .of Problems Xdentified Durin QA In-De th Ins ection The following items identified during the special quality con-trol inspection and reported in CO Report No. 244/69-1 have been resolved. The Applicant has been advised that the corrective measuxes or final resolution of the identified problems were con-sidered adequate by the inspector.

1. Weavin of Weld Beads in Excess of General Recommendations (Paragraph F. 3. b. (1) and (2)

Dr. Gilliland, consulting for Region X, reviewed at the Ginna site the weave-welding in stainless steel joints.

The nondestructive examinations of these joints were found to be acceptable. These results, together with the small percentage of stainless steel weave-welding were the bases for Dr. Gilliland's recommendation that the weave-welding be accepted, even though such practice is considered im-proper*. ~

• Parameter, Inc., Report No. DC-39, dated April ll, 1969, on file in Region I office.

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Messrs. Powell, Ford, Arthur, et. al., were informed that, the weave-welding at the Ginna Plant was accepted on a one case basis. The inspector stressed the following:

a. Acceptance on a one case basis should not be construed as condonement of improper welding techniques.

b. Weave-weldingis considered poor practice throughout industry.

Messrs. Ford and Powell stated that, efforts were being made by management in the Bechtel and Westinghouse organizations to eliminate all undesirable welding practices and techniques.

Messrs. Arthur and Powell stated that the welds on the two valves, manufactured by the Velan Mfg. Company, in the Residual Heat Removal System would be ground smooth and radiographs taken to insure that the welds were sound. Mr.

Powell said that weld preparation for radiography had not been done prior,to resolution of the weave-welding problem because he did not know what the consequences of grinding the welds would have been in relation to resolution of the problem.

2. Calibration of Tor ue Wrenches ara ra h F. 3. b. 4 Mr. Elliot stated that the torque wrenches had been sent to

'nspecial independent laboratory for calibration subsequent to the quality control inspection. The .independent labo-ratory performed detailed calibration measurements for each wrench, including the various size nuts required and 'orque the torquing values for each. The inspector reviewed the calibration curves generated and noted that the curves appeared to be within standard tolerances. The inspector stated that under the circumstances the values used for torquing, the nuts on equipment appeared,to be reasona'ble, but stressed the desirability of using a more reliable method to calibrate torque wrenches than checking one wrench against another.

3. Ex osure of Tendons to Moisture ara ra h F. 4. c. 1 6 2 Dr. Gilliland reviewed a test program conducted by Westinghouse to assess the extent of possible damage to the tendons subsequent to exposure to atmospheric corrosion and moisture. The protective grease (NO-OX-ID) is capable of absorbing up to 15% of water. Based on-the results of the test program and the moisture absorption characteristics of the NO-OX-ID grease, Dr. Gilliland determined-that the tendon integrity has not been impaired*.

The inspector stated that the Applicant's test program.to assess the extent and effects of tendon corrosion indicated that the integrity of the installed tendons has not been impaired. The results of tests conducted by Pittsburgh Testing Laboratory and a statement from W. R. Grace Company that the protective grease (NO-OX-ID) was capable of absorb-ing up to fifteen volume per cent water were the bases for acceptance.

Seven random samples were selected from, a tendon, that hpd been exposed to total weather for a period of approximately six months. These seven samples were tested by Pittsburgh Testing Laboratory. The test results show that the minimum ultimate tensile strength was observed to be 243,880 psi (Page 5.1.2-77 of the FSAR states that the minimum ultimate strength shall be 240,000 psi) and the ductility had not changed.

Mr. Arthur stated that the moisture had been removed. where possible and that in no case was protective grease pumped into a tendon casing until .almost all of =the free water had been physically removed.

The inspector stated that the exposure of the tendons and the tendon casings to atmospheric corrosion and moisture was in nonconformance with the specifications and. that careful lannin could have revented this undesirable condition.

• Parameter, Inc., Report,No. DC-39, dated April ll, 1969, on file in Region I office.

4. Diesel Generator Vaults ara ra h F. 5. c. l The inspector observed that an operable sump pump has been installed in each of the diesel generator vaults. The in-spector stated that the installation of sump pumps adequately corrected the possible diesel generator vault flooding prob-lem.

5. Sensor Lines Su l in Si nals ara ra h F. 5. c. 3 to the Safet In'ection S stem SIS The inspector observed that .the sensor lines supplying signals to the SIS from the pressurizer and the sensor lines supply-ing signals .to the auxiliary fe'edwater pump and the diesel generators from the steam headers have been partially pro-tected by metal coverings from incidents that could sever the lines. In addition, some of the sensor lines have been rerouted to reduce the potential loss of these lines. In the event that a loss of signal should occur, the desired actuation of the system. will take place. The inspector stated that the corrective action is adequate.

6. Redundant Circuits ara ra h F. 5. c. 4 The inspector observed that the output cables from, two of the three containment high pressure transducexs had been re-routed. The inspector stated. that, the previously observed deviation from the separation criteria of redundant circuits set. forth- in the FSAR had been corrected.

Mr. Arthur stated that a comprehensive. circuit tracing pro-gram* was in progress to insure that all, circuitry complies with the provisions of the,FSAR.

The inspector stated that the plant records would be audited and a possible spot check may be performed subsequent to the completion of the circuit tracing program to determine the effectiveness of the A licant's ro'am.

• CO Report No. 244/69-8, paragraph H.

J 'g Rela Room ara ra h F. 5. c. 5) inspector stated that a,fire in the relay room could

'he cause the loss of automakic control to,all systems. The in-spector acknowledged that the difficulties associated with this situation require a basic change in design philosophy and are not associated with any failure of the licensee to meet the requirements stated in the FSAR.

Mr. Arthur stated the fire fighting equipment in and near the relay room was the property of the erector and that RGsE fire fighting equipment would be installed at the earliest date possible. Mr. Mambretti said that the RG&E fire fighting equipment would be installed at the earliest date possible.

Mr. Mambretti said that the RGaE fire fighting equipment was in storage and that installation would be carried out simul-taneously with removal of the erector's equipment, following formal acceptance of the area by the Applicant.

Circuitr Test Resu1ts ara ra h F. 6. b. 2 The inspector stated that the Bechtel circuitry test results are deficient in that test failures and corrective actions are not recorded. The inspector audited circuitry test re-sults kept by the Applicant since receiving systems from Bechtel and noted that adequate records are being maintained.

These records include "trouble cards", equipment history records and log sheets. The inspector stated that the Applicant's records are adequate to maintain a history and equipmept,, status and maintenance record.

S ot Checkin of Installation b Westin house ara ra h F-. 6. b; 3 The Westinghouse Site Log (a proprietary record) was read by the inspector. This Site Log contains information that in-dicates spot-checking was carried out by several Westinghouse employees, including welding inspectors, engineers, manage-ment and other inspectors. The information- is not complete, hOwever, this Site Log does, reflect items found to be incon-sistent with Westinghouse Specifications'nd frequently in-dicates the type of workmanship being performed by the Erector's personnel. Results of followup action and final resolution of deficiencies are. recorded.,

Mr. Powell stated that the purpose of this Site Log was to outline problems and was not intended to show items that were considered to be satisfactory or did not require follow-up action.

10. Welder Control, Weldin Electrode Control and Weldin Fabrica-tion Records ara ra hs F. l. d. 2) a 3 and F. 2. d.

The welding problems identified by the special quality con-trol task group during the RG&E quality control inspection were reviewed by Dr. Gilliland*.

Dr.,Gilliland concluded that the most satisfactory method of obtaining assurance that improper welding rod was no/

used is to perform periodic nondestructive testing, volumet-ric as well as surface, during the early life of the plant.

This method of surveillance is based on, the conclusion that if improper welding electrodes were used, cracks can be developed after thermal cycling.

Boat sampling. or other sampling techniques involving removal of weld ance/or base material will give little assurance con-cerning the integrity af the welds.

All critical welds have been radiographed-.with no indication of defects. Chemical analysis of samples-appears to be the alternate method to identify potential defects. Inspection showed that a system existed for welding electrode control.

However, it was deficient in that some improper material could have been inadvertently used. An unrealistic number of samples would have to be analyzed to obtain meaningful data. Taking a large number of destructive samples with the potential problems which could be encountered in making the necessary repairs, appears to be unjustified.

• Parameter, Inc., Report No. DC-39, dated April ll, 1969, on file in Region I office.

Plant records and discussions with RGSE personnel revealed that all sj.te power was lost for a period of. approximately forty minutes on May 17, 1969. The cause of the outage was the trippi,ng of No. 6 transformer in the Ginna switchyard during a contrO1 board punching operation by electrical installation crew personnel. A second loss of power occurred during the inspector's visit .to the plant on May 20, 1969 while the plant was in a cooled down condi-tion. The second loss of power was also caused by a control board punching operation, Both diesel generators were out of service to investigate a pxoblem(discussed in paragraph D) when the outage occurred. Mr. Arthur stated that the Chief Electrical Engineer for RGGE had initiated new procedures for all maintenance, testing, installation and repair work to coordinate scheduled work to prevent recurrence of loss of site power or other abnormal electrical dis-tribution problems.

D.

Plant records and discussions with various RG&E personnel re-vealed that during a routine checkout of the diesels while the electrical bus was energized by normal electrical power, 'both diesels came on the line and attempted to supply power to the RG6E power distribution grid. Investigation showed that a relay had been installed .incorrectly in each system. This relay caused the "live" bus to indicate a "dead" bus. Fortunately, the diesel generators came into aervice in phase with the distribution grid and no damage to the equipment resulted. Both diesel'generator- units were secured manually when the output reached 2,000 KW. The diesel generators had never been tested under load conditions prior to this incident.

Both diesel generators were out. of service when an electrical outage occurred on May 17, 1969. The relays were installed correctly and tests will be performed to insure that, the diesel generator units function as designed.

E. Reactor Coolant Pum s RCP Plant records ynd discussions with various personnel indicate that No. 1-A RCP showed erratic cooling water flow to the labyrinth seals on May 19, 1969, and was secured. The RCP had. operated normally during all operating periods prior to this incident.

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10 The reactor coolant system was at, normal operating temperature, 540 F, -on May 17, 1969, when the constructiqn site experi,enced a loss of electrical power for approximately forty minutes (discussed in paragraph C of this report.) During the .electrical outage, no cooling water was supplied to either RCP.

Mr. Pow'ell stated that the seals on No. 1-A RCP would be replace/

with new seals and that an 'inspection would be made of the graphitar bearings to ipsure that, these bearings were in acceptable condition.

He said that if any indication of a possible bearing problem was discovered,- the bearings in both RCPs would be changed out pxior to continuing of Hot. Function Testing at the construction site.

F. Relief Valves

1. Main Coolant The inspector discussed the nondestructive testing (NDT) and acceptance criteria (AC)* for'he pressure retaining components of the safety valves with Mr. J. R; Zahorsky and

, reviewed carbon copies of records showing the NDT and AC performed on each valve. The 'original copy of each record is filed at the Wehtinghouse Document Center, pittsburgh, Pennsylvania. These records show the serial number of each valve and that, the following tests were performed:

a. -

Radiographs 100% of the valve disc and nozzle areas.

Results were satisfactory.

b. Dye Penetrant Testing the valve disc and nozzle areas.

Testing was performed following the final machining operation. No indication of any defects was ob'served.

c. Hydrostatic Testing - The valve discand nozzle were sub-jected, to 3110 psig '(design X1.25),'the body to 750 psig and the outlet side to 500 psig. Each hydro tept was observed for a period of 30 minutes. - No leakage was observed.

Seat Leakage , Tested at 2250 psig with steam, at the design set pressure of 2485 psig with air and for leak-age at 3110 psig while gagged. No indication of leakage was observed.

• CO Report No. 244/69-7, paragraph F. 2. a, and Memo from Mr. J. P, O'Reilly, dated 4/4/69.

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e. Operational Tests Operational .tests were performed on the valve spring, range of spring, ring setting, and the guide zing position was recorded.

f. Test Data The valve data included the xun number, opening pressure, closing pressure, blowdown pressure, back pressure and valve lift pxessure.

'and physical certifications were avai3,able.

Complete chemical The disc and seat on, each valve had-been haxd surfaced by pro-cedure.

The test, procedure,had been approved by Westinghouse, The valves and testing were in compliance with the re-c[uirements stated in BSPV Code,Section III.

2. Main Steam The valve testing for the main steam safety valises was per~

formed as described in paragraph l, except for the difference in pressures as specified for a design pressure of 1085 psig instead, of 2485 psig.

G. Valves in Safet Oriented S stems*

Mr. Mambzetti stated that a comprehensive'nspection of all

valves connected to the reactor coolant. and other safety oriented systems show that one valve contained carbon steel internals. These internals have been removed and stainless steel internals insta2,led.

There appears to be no other cax'bon steel intexnals in stainless steel valve bodies or carbon steel valves in any safety oxiented system.

H. Safet In 'ection S stem Pum s SIS Plant records and discussions with Messrs.-Sundstrom and Mambretti indicate that two of the three high pressure'IS pumps were operated'or a period of approximately four hours and then secured. Maintenance personnel attempted to rotate the shafts on these pumps by hand and determined that the shafts could not be turned. The shaft on the west um was turned with the aid of a

• Memo from Mr. J. P. O'Reilly, dated 4/8/69 and CY Report No. 68-12.

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12 stxap wrench. The shaft on the middle pump appeared to be frozen.

The Applicant requested technical assistance from the Worthington Pump Manufacturing Company, Harrison, New Jersey. Mr..Arthur stated that the 'SIS pumps would remain locked out until the Worthington representative and the Applicant determined and corrected the pump problem.

Mr. Arthur stated that he had telephoned management personnel at Niagara Mohawk POwer Company and was told that a similar ex- ~

pexience had occurred in a hydxaulic pump at that power station.

The inspector will continue'ollowup action on this item, and report the results of /he inspection. Particular attentiop will be directed towards the following:

1. Cleanliness of piping, strainers and. tanks upstream of these pumps.

2. Alignment of puris and motors.

3. Pipe sizing to .determine that adequate NPSH is available.

4. Potential design deficiencies which may have long term im-plications at Ginna or other plants.

I. Steam Generators SG The inspector observed that the code stamps, as required by the B&PV Code,Section III, had been attached to each SG.

Code inspectors had witnessed the required tests prj.or to of the code certifications and attachment of the 'ssuance code stamps.

2; Channel Head 7 Nut Plates

. The 7 nut plates* on the channel heady of -the SG have been radiograpQed, ultrasonically tested and heat treated at, 1150 F, according to Mr. Dolan. The inspection and furnace

'eat treat xecords are stored at the Westinghouse Document Center Tam a, Florida.

• CO Report, No. 244/69-6, paragraph G. 1.

e The stress relieving operation was conducted during field fabrication of the steam generators*.

J. Containment S stem

1. Containment Leak Rate Test Results "Gilbert Associates, Inc.-records, and discussion with Mr.

Sommers, indicated that the containment vessel leak rate**

test results were as follows:

a. Leak Rate 5 60 psig 0.0238% + 0.0168% per day or approximately 3.5 pounds/hr.

b. Leak Rate at 35 psig 0.0053% + 0.0180% per day or approximately 0.52 pounds/hr.

The design leak rate is 0.1% per day or less, Section 5, page 5.6.1-11 of the FSAR.

2. Air Circulatin -

S stem Mr. Arthur stated that several vanes in the air circulating system had broken off during the containment vessel over-pressurization and leak rate tests. The-vanqs were dis-covered subsequent to completion of the containment, vessel leak rate test, during equipment inspection, to determine damages. that occurred during the testing. The vanes were recovered and re-installed. Mr. Arthur stated that follow-up on the possible cause(s) for the loss of the vanes is being pursued.

K. New Fuel Ninety-six new- fuel elements had been received as of May 21, 1969.

The inspector observed that unloading, inspection, license rqquire-ments, safety precautions and security was. as described in CO Report No. 244 69-8 ara ra h I.

• CO Report No. 244/68-3, paragraph G.

• • CO Report No. 244/69-7, paragraph D.

L. Instrument Housin s Mr. Mambretti stated that a comprehensive inspection by RGGE personnel indicates that gaskets had been installed in all in-strumentation housings to preclude moisture from affecting the operation of instruments in the containment vessel.

M. Control Rod Drive. S stem Plant records and discussions with various RGSE and Westinghouse personnel indicate that the seal welds on the control rod drive pressure housings for the part length rods were hydrostatically tested at 2335 psig*.

N. H dro en Recombiner The inspector observed that, installation of the hydrogen re-combiners has started.

0. Instrument Air Su 1 Mr. Platt stated that plans have been completed to install a solenoid operated valve which will automatically bypass the pir dryer upon receipt of a supply header low pressure signal.

The inspector stated that this improvement may prevent a situa-tion similar to one that occurred at an operating power reactor**.

P. Hot Functional Testin Pro ram HFTd The inspector observed that less than 10% of the scheduled HFTP had been completed by May 21, 1969. Delays in the HFTP arq the re~

suit of a work stoppage by the Rochester Local of the Insu1ator~s Union and a problem with the labyrinth seals on No. 1-A Reactor Coolant Pump.

Q. Fuel Loadin Date A revised fuel lqading date was unyvailable. The resolutiqn of the Reactor Coolant Pump seal problem and a work stoppage by the Insulator's Union have resulted in a significant loss of time. A revised fuel loading date will be reported by the inspector at the earliest date ossible.

• CO Report No. 244/69-8,,paxagraph C. 1.

~*CO Report No. 29/67-4, paragraph I.

V 15 R. Tendon Ins ection The inspector observed that preparations were in progress for the 1,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> inspection of the stressed tendons*. Results of this inspection will be obtained and reported at a future date.

S. Exit Interview An exit interView was held with Messrs. Arthur, Platt, Sundstrom and Mambretti. The principal subjects discussed were as follows:

1. Problems identified during the QA In-Depth Inspection.

Significant comments are included in the appropriate para-graphs of the report.

2. Power Outages. Mr. Arthur stated that the loss of power indicated that improvement in communications and work scheduling were necessary and that steps had been taken to alleviate this type of problem.

3. Diesel generators phasing into the RGSE distribution grip.

Mr. Arthur stated that the diesel generators had not been tested under load conditions and that serious damage to both diesel generator units could have occurred.

4. Reactor Coolant Pump Seal Problem. Comments included in the appropriate paragraphs of this report.

5. Relief valves and valves in safety oriented systems. The inspector stated that the information supplied by Crosby Valve Company representatives indicated that the required tests had'een performed. Mr. Arthur stated that RGSE was satisfied with the information and installation of these valves.

6. .

Pumps and Motors. Mr. Arthur stated that the splash barriers for the Residual Heat Removal Pump Motors should be adequate protection to prevent wetting the motors. He stated tpat the Reactor Coolant and Safety Injection pumps would be repaired and tested to insure trouble free opera-tion.

• CO Report, No. 244/69-8, paragraph F.

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7. Hot Functional Testing Program. Mr. Arthur stated that the hot functional testing was behind schedule due to the In-sulator's Strike and the reactor coolant pump labyrinth seal problem.

8. Fuel Loading Date. Mr. Arthur stated that no revised fuel loading date would be available until the results of the reactor coolant pump seal inspection and settlement of the Insulator's Strike were known.

9. Sign off of welder control, welding electrode control and welding fabrication records problems identified during the QA In-Depth Inspection. The inspector informed Mr. Arthur during telephone conversation, subsequent to completion of the inspection, that these problems would be signed off and that a surveillance. program would be included in the Tech-nical Specifications to assure that improper welding rod had not been used in critical welds.

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