ML20202J769
| ML20202J769 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Browns Ferry |
| Issue date: | 03/18/1986 |
| From: | Henrich N, Slagle F, Stevens W TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML082340316 | List: |
| References | |
| RTR-REGGD-01.097, RTR-REGGD-1.097 I-85-773-BFN, NUDOCS 8604160175 | |
| Download: ML20202J769 (7) | |
Text
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TENNESSEE VALLEY AUTHORITY NUCLEAR SAFETY REVIEW STAFF NSRS INVESTIGATION REPORT NO. I-85-773-BFN EMPLOYEE CONCERN:
XX-85-122-026
SUBJECT:
THERMAL OVERLOAD BYPASS AND INDICATION PROBLEMS DATES OF INVESTIGATION:
MARCH 1-5, 1986 3!/[
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'tw INVESTIGATOR:
N. T. HENRICH DATE b ' d -I k ).
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REVIEWED BY:
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DATE F. J. SLAGLE APPROVED BY:
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W. D. STEVENS DATE l
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8604160175 860411 PDR ADOCK 05000259 P
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BACKCROUND A Nuclear Safety Revieu Staff (NSRS) investigation was conducted to determine the validity of an expressed employee concern received by the Quality Technology Company (QTC)/ Employee Response Team (ERT). The concern of record, as summarized on the Employee Concern Assignment Request Form from QTC and identified as XX-85-122-026, stated:
Browns Ferry: Thermal overload bypass and indication problems involving Reg. Guide 1.97.
CI has no further information. Anonymous concern via letter.
Further information was requested from the ERT follow-up group to identify specific problems referred to in the concern of record. No additional information was available.
II.
SCOPE A.
The scope of this investigation was determined from the stated concern of record to be that of a single, specific issue requiring investigation.
8 Thermal overload bypass circuits used on motor operated valves (MOV's) may not meet the require-ments of NRC Regulatory Cuides.
B.
To accomplish this investigation, NSRS reviewed applicable NEC regulatory guides and industry standards to determine the require-ments governing the use of thermal overloads on safety-related, motor-operated valves. The use of thermal overloads for motor protection on Browns Ferry Nuclear Plant's (BFN) critical structures, systems, and components (CSSC) motor-operated valves l
was determined from as-constructed drawings for all three units and compared to the stated requirements.
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III.
SUMMARY
OF FINDINGS A.
Requirements and Commitments 1.
U.S. NRC Regulatory Cuide 1.106, " Thermal Overload Protection for Electric Motors On Motor-Operated Valves" (Ref. 1).
2.
IEEE Standard 279-1971, " Criteria for P'rotection Systems for Nuclear Power Generating Stations" (Ref. 2).
B.
Findings 1.
Regulatory Cuide 1.106 (Ref. 1) specifies that thermal overload protection devices which are integral with the motor starter for electric motors on motor-operated valves (MOVs) 1
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t se e o i e
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shall not needlessly provent the motor from performing it1, safety-related function. The thermal overload protection devices must not remove power from safety-related MOVs before the valve's safety function has been completed.
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2.
Regulatory Guide 1.106 (Ref. 1) defines two alternatives to ensure the requirements of paragraph III.B.1 are met.
They are as follows:
a.
Regulatory Position 1 1
(1) Thermal overload protection devices may be continuously,.
bypassed and temporarily placed in service when the valve motors are undergoing periodic testing or maintenance.
i (2) Thermal overload protection devices may ba continuously 1
in service during normal plant operation and bypassed l
under accident conditions. The bypass initiation l
system circuitry should conform to the requirements of IEEE standard 279-1971 (Ref. 2), sections 4.1, 4.2, 4.3, 4.4, 4.5, 4.10, and 4.13.
I b.
Regulatory Position 2 i
d The thermal overload protection device trip _setpoint may be established with uncertainties in setpoint drift, inaccu-racies in motor heating data and ambient temperatures at the installed location of the valve motor, and its overload protection device resolved in favor of the MOV completing its safety-related function.
3.
BFN is not committed to Regulatory Guide 1.106.
NOVs whose operation is vital to reactor safety are equipped, however, with thermal overload protection devices selected to permit valve motor operation at locked rotor current for 16 to 30 seconds.
This ensures that the valve motor is not tripped prematurely and that sufficient time is given for the valve to complete its 3
required travel. This is consistent with Regulatory ?osition C.2 of reference 1.
4.
Thermal overloads were initially set in accordance with guidelines set forth in reference 5.
These overloads are not addressed in plant technical specifications and are not required i
to be periodically tested.
5.
BFN Electrical Maintenance Section is presently surveying all 480-volt MOV boards, recording the type of heater element utilized on the MOVs and the thermal overload setting.
BFN j
personnel have indicated that this data will be forwarded to the-l Office of Engineering (OE) for review.
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A teview of as-constructed drawings listed in reference 6 confirmed that thermal overloads remain in the MOV motor control circuit at all times.
7.
BFN does not utilize bypass circuits for thermal overloads on safety-related MOVs.
However, BFN engineering change notice ECN L2071 (Ref. 7) will add bypass contacts to MOV control circuits j
to bypass thermal overloads on critical valves in the event of l
an accident.
This ECN is scheduled for unit 3 cycle 5 outage but is not
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currently scheduled to be worked for units 1 or 2.
8.
In April 1983, Black and Veatch submitted a report of their independent review of the Watts Bar Nuclear Plant (WBN)
Auxiliary Feedwater System to TVA.
Findings F108 and F122 of this review questioned the adequacy of the WBN design of the MOV thermal overload relay bypass circuitry and bypass indication relative to Regulatory Guide 1.106 (Ref. 1).
These findings are not applicable to BFN because of the differences in the thermal overload relay heater circuitry designs between BFN and WBN.
This was confirmed by the TVA Task Force for review of Black and Veatch findings.
9.
Regulatory Guides 1.97 (Ref. 3) and 1.47 (Ref. 4) are not applicable to the concern of record.
IV.
CONCLUSIONS AND RECOMMENDATIONS
.A.
Conclusions The concern of record was not substantiated by this investigation because the thermal overload protection devices for valve motors on safety-related MOVs at BFN are not bypassed during accident conditions.
The design of the MOV motor control circuits conform to Regulatory Position 2 of Regulatory Guide 1.106.. Engineering Design provided guidelines for the initial setting of the thermal overloads on critical MOVs. However, these thermal overloads are not required to be tested periodically by plant technical specifications to verify their settings.
B.
Recommendations 1.
I-85-773-BFN-01, Survey of 480 Volt MOV Boards Provide OE with the results of the current survey of 480 volt MOV boards for review of the thermal overload settings.
2.
I-85-773-BFN-02, Overload Trio Settinzs Guidelines should be issued to clearly define the thermal overload trip settings on critical M07s, and the overloads settings verified periodically.
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DOCUMENTS REVIEWED IN INVESTICATION I-85-773-BFN AND REFERENCES 1.
U.S. NRC Regulatory Guide 1.106, " Thermal Overload Protection for Electric Motors on Motor-Operated Valves," Revision 1, dated March 1977
)
2.
IEEE Standard 279-1971, " Criteria for Protection Systems for Nuclear Power Generating Stations," approved June 3, 1971 3.
U.S. NRC Regulatory Guide 1.97. " Instrumentation for Light-Water-Cooled, Nuclear Power Plants to Assess Plant and Environs Condition During and Following an Accident," Revision 3, dated May 1983 4.
U.S. NRC Regulatory Guide 1.47, " Bypassed and Inoperable Status Indication for Nuclear Power Plant Safety Systems," dated May 1973 5.
Memorandum from J. R. Parrish to W. P. Kelleghan, " Browns Ferry Nuclear Plant Units 1-3 Preoperational Test Procedures for Critical Motor-Operated Valves," dated May 19, 1971 6.
TVA As-Constructed Drawings a.
45N779 Drawing Series Sheet Revision Sheet Revision 6
RF 27 RB 8
RC 28 RA 9
RG 29 RA 10 RG 30 RA 11 RF 31 RA 12 RF 32 RB 13 RA 33 RB 14 RA 34 RB 15 RA 35 RB 16 RA 36 R1*
17 RI 37 R2*
18 RA 38 R2*
19 R2*
21 RF 41 RB 22 RE 42 RO 23 RG 43 RC 24 RB 25 RB i
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45N714 Drawing Series Sheet Revision 2
RB 3
RA 4
R7*
5 RB 6
RA 7
RA 8
R1*
9 22*
10 R1*
- As designed drawings 7.
BFN Engineering Change Notice (ECN) L2071, " Modification to Motor-Operated Vaive Control Circuits," dated August 22, 1977 e
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UNITED STATES GOVERNMENT Memorandum TENNESSEE VALLEY AUTHORITY TO M. L. Abercrombie, Site Director, Sequoyah Nuclear Plant FROM K. W. Whitt, Director of Nuclear Safety Review Staff. E3A8 C-K DATE
- March 28, 1986
SUBJECT:
NUCLEAR SAFETY REVIEW STAFF INVESTIGATION REPORT REVISED TRANSMITTAL This memorandun supersedes K. W. Whitt's memorandum dated March 17, 1986 on the same subject and is being issued to clarify responsibilities for responding to the recomendations.
Transmitted herein is NSRS Report No. I-85-861-SQN.
Subject:
GENERIC IMPLICATIONS OF WATTS BAR KER0 TEST VALVE PROBLEMS ON SQN Concern No.: II-85-090-002 and associated prioritized recomendations for your action / disposition.
This report contains two Priority 1 [P1] recomendations which must be addressed before startup. The report also contains one Priority 2 [P2]
recomendation concerning Watts Bar Nuclear Plant. The Watts Bar Employee Concern Task Group should include this recomendation in their handling of the WBN employee concern processing.
Should you have any questions, please contact W. D. Stevens at extension 6231-K.
'Recomend Reportability Determination: Yes No X
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1 BFS:BRP i
c. e ! ;1 Attachment cc (Attachment):
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I[_-(('i_i)-t-j "iid,j W. C. Bibb, BFN W. T. Cottle. WBN J. P. Darling, BLN 20,3-i i-P '
R. P. Denise, LP6N40A-C (P2 recommendation for yotir%sh'dl'iriiDZl.~.'
G. B. Kirk, SQN
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~'*-Q,,j N. L. Martin, WBN
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QTC/ERT, WBN
-..--.--i E. K. Sliger, LP6N48A-C J. H. Sullivan, SQN (2) 2 0616U
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Rue TT C Ce,.;ene Rnerle Roaulariv nn the Paventi.taninos Plan s~.
TENNESSEE VALLEY AUTHORITY NUCLEAR SAFETY REVIEW STAFF NSRS INVESTIGATION REPORT NG. I-85-861-SQN EMPLOYEE CONCERN:
XX-85-090-002 INTERIM REPORT
SUBJECT:
GENERIC IMPLICATIONS OF WATTS BAR KER0 TEST VALVE PROBLEMS ON SEQUOYAH DATES OF INVESTIGATION:
DECEMBER 5, 1985 - MARCH 6, 1986 3 9 [d INVESTIGATOR:
C. L. BREEDING '
DATE /
as INVESTIGATOR:
[M-M 3//7 J. C. CATLIN DATE REVIEWED BY:
[b 3//7/(,
L. E. BROCK DATE APPROVED BY:
obdW 3//7/N
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W. D. STEVEN 3
' DATE '
' po#&7-0-380"
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BACKGROUND A Nuclear Safety Review Staff (NSRS) investigation was conducted to determine the validity of an expressed employee concern received by Quality Technology Company (QTC)/ Employee Response Team (ERT). The concern of record, as summarized on the Employee Concern Assignment Request Form from QTC and identified as XX-85-090-002, stated:
Sequoyah: Unit l' & 2.
Per CI TVA used globe valves (Kerotest) extensively in both plants, Watts Bar and Bellefonte and had leakage & corecsion problems. CI questions the usage of these valves at Sequoyah - the sister plant - for leakage & corrosion problems. The systems to be checked as examples are CVCS, safety injection, RHR & reactor coolant, etc. CI has no further information. NUC POWER concern.
A similar concern has been raised for Kerotest valves installed at Bellefonte Nuclear Plant (BLN). This concern will be addressed by the investigation of concern KX-85-090-001.
II.
SCOPE A.
The scope of this investigation was determined from the" stated concern of record to be that of two specific issues requiring investigation:
1.
Verification of whether Watts Bar Nuclear Plant (WBN) has experienced extensive leakage and corrosion problems with Kerotest globe valves.
2.
Determination of any safety implications for Sequoyah Nuclear Plant (SQN).
B.
The failure of these valves at WBN was investigated through document reviews and personnel interviews. The specific use of Kerotest valves at SQN was investigated.
The generic implications of valve failures at WBN upon the safe operation of SQN were also investigated, and industry experience with these valves was reviewed through the Nuclear Plant Reliability Data System
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(NPRDS).
C.
This report will be issued as an interim repart. Revision 1 will be issued when the results of further investigation of supporting documentation become available. Missing documentation-includes Westinghouse specification C-678824 Rev. 7 and TVA specifications applicable to WBN purchases.
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III.
SUMMARY
OF FINDINGS e
A.
Requirements and Commitments 1.
10 CFR 50.55(e) of the Title 10 Code of Federal Regulations (January 1, 1985 Edition) (Ref. 1) requires that the holder of a construction permit for a nuclear plant notify the NRC of each deficiency found in design and construction which, were it to have remained uncorrected, could have affected adversely the safety of operations of the nuclear power plant at any time "
throughout the expected lifetime of the plant.
2.
10 CFR 21, " Reporting of Defects and Noncompliance," of the Title 10 code of Federal Regulations (January 1, 1985 Edition)
(Ref. 2) requires reporting to the NRC of substantial safety hazards or that components supplied to a nuclear plant contain defects which could create a substantial safety hazard.
3.
TVA Specification 9923 (no date - no revision number),
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" Principal Piping Systems and Appurtenances - Sequoyah Nuclear i
Plant Units 1 and 2," was attached to contract 71C37-92615.
j This contract was to NAVC0 for piping and valves for SQN.
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B.
Findings i
j 1.
A problem with Kerotest valves at WBN was identified and documented in a Division of Construction Nonconformance Report (NCR) 2501R dated October 20, 1980 (Ref. 4).
Ihe final report on this problem was transmitted from John A. Raulston to L. M.
Mills on April 27, 1981 (Ref. 3).
In that report the
" Description of the Deficiency" identified several hundred 3/4, 1, and 2-inch valves with leakage and corrosion j
problems. The " Safety Implications" section of the report states:
While some of the subjec6 Kerotest valves are installed in essential safety-related systems:
CVCS, SIS, RHR, UHI, RCS, and CSS, operation of the valves is not required for the safe shut-down of the plant during a loss of coolant accident. As a result, TVA could identify no valves that perform a safety function.
However, to document this, a failure effect i
analysis was performed. The analysis identified no detrimental effect on plant j
safety as a result of the failure of any of these valves.
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The report also stated in the " Corrective Action" section:
TVA does not consider the valves to perform a safety function; however we do believe that the corrosion identified with the valves could result in a maintenance problem during the life of the plant. Therefore, TVA has instituted a maintenance program to dismantle, inspect, and replace parts as required for those valves installed at WBN.
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In addition, the report noted the " generic applicability of the corrosion problem" and stated:
Verbal discussions with personnel at GQN indicated that during normal disassembly for maintenance, they have identified no Kerotest valves with what they consider excessive corrosion.
A letter from L. M. Mills to James P. O'Reilly dated April 24, 1981 (Ref. 11) transmitted this report to NRC in accordance with the requirements of 10 CFR 50.55(e) and 10 CFR 21,.
However, no mention was made that the WBN procurement specification was deficient by not requiring valve packing removal and bonnet drying af ter hydrostatic testing.
2.
The response to NCR 2501R does not address the safety consideration of a 1-percent fuel failure as stated in the safety analysis. If their stems leaked, are these valves in areas where they might impair personnel entry which in turn might be required to maintain other safety functions?
3.
On September 10, 1981, R. W. Cantrel'1 wrote a memo to J. A. Raulston on the subject of "Saquoyah Nuclear Plant Units 1 and 2 - Deficient Kerotest Y-Type Globe Valves - Report No. 4 (Final)" (Ref. 8).
This memo was in response to the commitment-made in NCR 2501 to follow up on generic implications of the Kerotest valve failures. This memo stated:
We have reviewed the Korotest valve installa-
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tions at Sequoyah Nuclear Plant and find that these valves' safety function would not be compromised by problems daveloping from a water saturated stem packing. Westinghouse provided the majority of the valves installed in safety systems of SQN. Westinghouse specified packing replacement after hydro-static testing on their procurements.
4.
The Westinghouse specification applicable to these valves is G-678824, Rev. 7.
NSRS has been unable to obtain a copy to verify the quoted statement above as of.the date of this report.
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b" The Westinghouse-procured valves were dedicated to the Nuclear Steam Supply System (NSSS). Most of them were designed with a unique valve body length.
5.
The remainder of the small valves for SQN were procured either by NAVC0 as part of the principal piping contract (71C37-92615) or directly by TVA on contract 824147. TVA specifica-tion 9923 (no revision number, no date) was used in these purchases. This specification did not require that valve bonnets be dried out or packing removed after hydrostatic tests.
6.
The response following review of SQN Kerotest valves (Ref. 8) stated that most of the valves at SQN were procured by Westinghouse and implied that there was no problem at SQN.
7.
Most of the Kerotest valves used at WBN were also procured through the Westinghouse NSSS contract. The remainder were procured directly by TVA.
Applicable specifications and requirements for WBN remain to be investigated.
8.
A search was made of Maintenance Requests (MRs) filed at SQN since the plant went into operation. The object was to find valves that had failed or needed repair due to leakage or corrosion; 128 failures were found in the files. None of these failures were Kerotest Y-type globe valves.
9.
A search of the NPRDS (a nationwide data base for operating nuclear plants) revealed operating data on over 1600 Kerotest globe valves in their data base. Of these valves, 60 failure reports have been filed with the system over the past ten years. Of those 60 failures, only four were caused by corrosion. These failures were reported from non-TVA nuclear plants. SQN participates in the NPRDS; WBN (since it does not have an operating license) does not. None of the Kerotest valves included in the system for SQN have failures reported.
10.
Maintenance personnel at SQN were interviewed to obtain information on Kerotest valve maintenance.
They were aware of the WBN problems, but they could recall only two problems with i
Kerotest valves. These problems were the results of improper-
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installation of the yoke of the valves and not related to the corrosion and leakage problem.
11.
A search of TVA's Equipment Identification System (a comput-erized data base containing data on plant equipment) for Kerotest valves installed at SQN was made; 1,528 Kerotest globe valves are listed as installed at SQN. The size of all of these valves is two inches or less. A second search for non-Westinghouse valves found over 500 Kerotest globe valves installed in SQN CSSC systems that were not provided by Westinghouse. These valves are used extensively in the ERCW system and as instrument isolation valves, root valves, sample connections, vent valves, and drain valves for other systems.
4
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Space Kerotest Y-type globe valves were found in the ECN warehouse at SQN. A one-inch Kerotest valve was located in i
the ECN warehouse, and a workplan was written to disassemble the valve.
The manufacturar's identification contained the following information:
serial number, SY4-1; class, 2; year built, 1975; size, 1"; TVA-D-9916; 39875-5198. A second tag attached to the valve had two numbers on it which read:
P. O. 39875-5198 and 47W495-303. The valve was disassembled by SQN mechanical maintenance at the request of NSRS. A clear liquid that looked like water was in the bonnet area. The bearing on the stem was corroded. Rust was visible on parts
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inside the bonnet. The bearing would rotate, but it was not -
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smooth. Photographs were taken of the valve parts and will be retained in the NSRS files.
13.
Three former TVA field inspectors were interviewed. The consensus was that packings were generally removed at the various suppliers because it was good practice and not because it was stated in the specifications. The asbestos / graphite packing generally used during the time period when these valves were shipped would retain water; and, hence, could cause corrosion and/or seizure of the valve stem. Source inspection of smaller valves was sometimes waived makips their*
condition uncertain.
V.
CONCLUSIONS AND RECOMMENDATIONS A.
Conclusions 1.
WBN has experienced leakage and corrosion problems with Kerotest valves, and the resolution of this problem is documented in NCR 2501.
Over 500 Kerotest globe valves purchased on non-Westinghouse 2.
contracts are installed in CSSC systems at SQN. These valves appear to have been subjected to the same type of factory hydrostatic test as the WBN valves.
3.
Inspection of a one-inch Kerotest valve by this investigation revealed corrosion on the valve bonnet internals. This investigation could find no documentation that indicated that the Westinghouse hydrostatic testing procedure was better than that in the TVA specification. Thus, there could be concern about the ability of the approximately 1,000 Kerotest valves procured by Westinghouse being able to function for their
.l expected 40-year life.,
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4.
The Kerotest valves at SQN have been in service for many years and have no history of problens. Also, tests at the Kerotest factory (Ref. 10) have shown that these valves can operate even though corrosion is present in the stem area. In addition, safety evaluations have been performed that indicate even if the SQN Kerotest valves were subject to extensive corrosion and leakage, their safety function would not be impaired.
5.
The safety issue of the secondary effect of stem leakage during '
a 1-percent fuel failure accident was not addressed.
6.
Although there may be no compromise of the " safety function" 'of the Kerotest Y-type globe valves themselves at SQN by their failure, corrosion in the stem area would result in mainte-nance problems during the remaining life of the plant. Leaking i
valves could result in plant inaccessibility, inoperability, and excessive personnel radiation exposure.
7.
The employee concern is substantiated because there has been no l
objective attempt to evaluate this generic concern about the
]
Kerotest valves at SQN.
B.
Recommendations 1.
I-85-861-SQN-01, Kerotest Valve Inspection i
Although there may be no compromise of the " safety function" of the Kerotest Y-type globe valves at SQN by their failure, I
corrosion in the stem area would result in maintenance problems during the remaining life of the plant and leaking valves could l
result in plant accessibility, operability, and excessive j
personnel radiation exposure.
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Therefore, an inspection should be made of representative i
Kerotest valves at SQN.
If corrosion is found in the stem i
area, or stem leakage is found,-an engineering evaluation l
should be performed to determine the reportability and proper.
resolution of this problem.
If corrosion and leakage are not j
found, an evaluation should be performed to determine why the difference exists between WBN and SQN valves.
i OE should evaluate the methods used to determine generic applicability and not rely solely on verbal information such as that received from SQN.
[P1).
2.
I-85-861-SQN-02, NRC Reportability j-WBN should provide objective evidence that the statements regarding the Westinghouse testing procedure contained requirements for drying out bonnets and/or replacement of i
bonnet packing after hydrotesting were actually in the applicable Westinghouse specification.
Facts to the contrary should be reported to the NRC.
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WBN should notify the NRC in accordance with 10 CFR 50 Part 21 of the root caust of the WBN valve failures.
This is presumed by the NSRS to be lack of proper precautions taken after hydrostatic testing. [P2]
3.
I-85-8(1-SQN-03, Analyze 1-Percent Fuel Failure Accident The impact of Kerotest valve stem leakage coupled with the 1-percent fuel failure accident stated in the safety analysis should be investigated. [P1]
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DOCUMENTS REVIEWED IN INVESTIGATION I-85-861-SQN AND REFERENCES 1.
10 CFR 50.55(e) 2.
10 CFR 21, " Reporting of Defects and Noncompliance," dated January 1, 1985 3.
Memorandum from John A. Raulston, Chief Nuclear Engineer, to L. M.
Mills, Manager, Nuclear Regulation and Safety, " Watts Bar Nuclear-Plant Units 1 and 2 - Deficient Kerotest Y-Type Globe Valves -
Report No. 4 (Final) - NCR 2501R," dated April 27, 1981 (810430C0153) 4.
NCR 2510R R1 dated October 22, 1980 (801027B0425) 5.
NCR 2510R dated August 4, 1980 (8008270D0133) 6.
NCR 2272R dated April 18, 1980 (WBN 8004256003) l 7.
Memorandum from R. M. Pierce, Sequoyah and Watts Bar Design Projects Manager, to J. E. Wilkins, Project Manager, Watts Bar Nucidar Plant, " Watts Bar Nuclear Plant - Nonconformance Report No.
21272R," dated May 12, 1980 (SWP 800513 009) 8.
Memorandum from R. W. Cantrell, Sequofah and Watts Bar Design Projects Manager, to J. A. Raulston, Chief Nuclear Engineer, "Sequoyah Nuclear Plant Units 1 and 2 - Deficient Kerotest Y-Type Clobe Valves - Report No. 4 (Final)," dated September 10, 1981 (810916F0144) 9.
Letter from C. J. Transue, National Sales Manager, Kerotest Manufacturing Corporation, to Larry Tummel, TVA, dated May 29, 1980 10.
Letter from C. J. Transue, National Salen Manager, Kerotest Manufacturing Corporation, to Larry Tummel. TVA, dated August 11, 1980 11.
Letter from L. M. Mills, Manager Nuclear Regulation and Safety, to
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James P. O'Reilly, Directcr Office of Inspection and Enforcement, NRC, " Watts Bar Nuclear Plant Units 1 and 2 - Deficient Kerotest Y-Type Globe Valves - NCR 2501R - Final Report," dated April 24, 1981 (A27 810424 014) 12.
TVA Design Specification, WBN-DS-1935-1521-CK, R2, dated August 20, 1976,
" Motor-Operated and Manual Valves" 13.
TVA Division of Engineering Design, WB-DC-40-31.2, Watts Bar Nuclear Plant, Design Criteria for Seismic Qualification of Category I Fluid System Components and Electrical or Mechanical Equipment, R1 dated October 25, 1974 8
DOCUMENTS REVIEWED IN INVESTICATION I-85-861-SQN AND REFERENCES (Continued) 14.
Contract 71C 37-92615 between TVA and National Valve & Mfg. Co. (NAVCO) dated August 26, 1970 15.
TVA Specification 9923 for Principle Piping Systems and Appurtenances Sequoyah Nuclear Plant Units 1 and 2 (attached to contract 71C 37-92615) no date, no revision number 16.
Westinghouse Equipment Specification G-678824, "2 Inch and Belew Manual Valves (Class 1, 2, and 3 of ASME Boiler and Pressure Vessel Code,Section III)," Rev. 1, dated December 9, 1975 17.
Westinghouse Purchase Orders 106765, 146831, 178250, 178257, and 178258 under TVA contract 68C60-91934 18.
TVA Specification MEB-SS-10.19, Rev. O, dated November 17, 1977,
" Technical Specification for ASME Code Valves, 2 Inches and Smaller" f
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UNITED STATES, GOVERN 3 TENT s
V Memorandum Tsxxsssss vAttsy Aurir"qo,r iry W. C. Bibb, Site Director, Browns Ferry Nuclear Plant TO R. B. Kelley, Director of Nuclear Quclity Assurance, LP4N45A-C K. W. Whitt, Director of Nuclear Safety Review Staff,-E3A8 C-K FROM DATE
- April 2, 1986 SUJJE',T : NUCLEAR SAFETY REVIEW STAFF INVESTICATION REPORT TRANS'!ITTAL a
Trans.'itted herein is NSRS Report No.
_ I-85-516-BFN Subject IMPROPER TRACKING AND DISPOSITIONING OF CERTAIN
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NONCONFORMANCE REFORTS Concern No.
-516-BFN and associated recontnendations for your action / disposition.
_ It is requested that you respond to this report and the attached reco:=nendations by May 2,1986.
Should you have any questions, please contact M. A. Harrison at extension 6328-K.
Reco mend Reportability Determination: Yes No X j'/
N K. W. Whitt
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I MAH:BRP Attachment cc (Attachment):
M<L.Myh J N
- 11. L. Abercrombie, SQN 3
W. T. Cottic, WBN
.D. R. Nichols, E10A14 C-K J. P. Darling, BLN QTC/ERT, WBN I
.R.
P. Denise, LP6N40A-C E. K. Sliger, LP6N48A-C
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--Copy and Return--
To :
K. W. Whitt, Director of Nuclear Safety Review Staff, E3A8'C-K l
From:
l Date:
l, j-I hereby acknowledge receipt of NSRS Report No.
Subject for action / disposition.
i
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Signature Date i
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