ML20150D318
| ML20150D318 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 03/11/1988 |
| From: | NRC OFFICE OF SPECIAL PROJECTS |
| To: | |
| Shared Package | |
| ML20127A683 | List:
|
| References | |
| NUDOCS 8803230312 | |
| Download: ML20150D318 (22) | |
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[,-%D UNITED STATES
'i NUCLEAR REGULATORY COMMISSION
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SEOUOYAH NUCLEAR POWER PLANT, UNITS 1 AND 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERNS ELEMENT REPORT MC-40703-SCN "HEAT CODE AS RELATED TO MATERIAL C0tiTROL" I.
SUBJECT Category:
Materials Control (40000)
Subcategory:
Procedural Control Elenant:
"Heat Code as Related to Material Control" (40703)
The basis for elcrent report MC-40703-SCN, Revision 2, dated May 12, 1987 is employee concerns IN-85-545-X07, W1-85-008-C02, XX-85-027-X02, EX-85-023-001 and IN-85-660-001.
Three of the concerns, IN-85-545-X07, WI-85-008-002, and XX-85-027-X02, related to a lack of credibility of mthods used during construction to be certain that properly certified materials have been installed during construction, The other *.wo concerns, r.X-85-023-001 and IN-85-66C-C01, related to a lack of credibility of mthods used during plant rodification performed af ter the plant was placed in operation.
II.
SIM 9RY OF ISSUES The following issues were defined by TVA:
A.
Construction _
The perceived problem, as corived frcm concerns IN-85-545-XO7, WI-85-008-002 and XX-85-027-X02 is that there is a lack of credibility of methods used in the Construction Program, Heat Nurter Sort Printcut (HNSP) for verification of properly certified Pressure Boundary Materials, et installation.
B.
suclearPower The perceived problen as derived from concerns EX-85-023-001 and IN-85-660-001 is that there is a lack of credibility of methods used in the Nuclear Pwer Program Power Storerecm Requisition (Form TVA-575), for verification c 3 p'operly Certified Pressure Scundary Material at installation.
'II. EVALUATION A.
TVA's Review Sumary TVA designated an Erployee Concern Ta!.k Group on July 1, 1986 te investicate these ccncernt.
The results of this investigation war
- documented in TVA Elen nt Peport No, MC-40703-SON.
This report identified mora than 200 cossit'c discrepancies between Secucyah Units 1 and 2 cn safety-related piping (9? et Unit 1.:nd 110 at Unit 2).
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. The following corrective actions have been implenented to correct the existing problerns identified by the ECTG Report and to preclude their recurrence:
1.
PIRSONNEBS638 will ensure the clear definition of the applicable code edition and addenda of ANSI B31.7 used in the fabrication, erection, installation, and use of Nuclear Class Piping components, in the upper-tier documents.
(CATO No. 40703-SCN-01-R2 and CATO No.
40703-SGN-03-R0.)
2.
CACR SQP870627 will ensure that all Nuclear Class I, II, and III (TVA Class A, B, and C/D) pressure-retaining piping ccmponents will be examined and their suitability for use verified and docunented in accordance with the applicable recuirements, or replaced.
(CATD No.
4C703-SCN-02 90, CATO No. 40703-SCN-06-R0 and CATD No. 40703-SON-07-RO.)
3.
CAR-36-C64 will ensure that site procedures contain the necessary detailed instruction to provide for the receipt, storage, and installatien cf Nuclear Class Piping Cottponents in ccepliance with the applicable code recuirements.
(CATO No. 40703-SON-04-RO.)
4 CAR-84-064 will ensure that inspectors will receive the required training to ensure that Nuclear Class Piping Cceponent raterial identification verificaticn is perforced and docutanted, in rictordance with the 1
applicable code recuirements, throughout their receipt, storage, end installatien at SON.
(CATD No. 40703-SQN-05-RO.)
5.
SCRSONME98614 R1 and ECN L6784 will ensure that TVA design drawings contain cicar and consistent identification of where (location) and hcw (e.g., double autoratic valve, specially bored fitting) the piping classification changes, as stated in the FSAR, are effected.
(CATO ho.
40703-SCN-OS-RO.)
6.
FIRSGhMEB8793 will cr.sure that either the FSAR or the design drawing contain a clear definition of the bcundary betwcen the prirary coolart loops and their branch lines.
(CATO ho. 40703-SCN-09-R0.)
TVA (Division of Nuclear Enginecting) then us mbled a new investigative unit, tiie Feat Code Traceabil:ty Task Group (HCTTG) to evaluate and resolve the issues raised by the ECTG. The results of this investigaticn were docutented in TVA's report 825870225-C36.
This report (B25370225-036) reduced the 208 original discrepancies to a total cf 7 items of ncnccmpliance.
Tha investigations led to the issuance of three Corrective laction Reports 1
(CARS)--SQ-CAR-86-052, 50-CAR-86-C55, and 50-CAR-26-064--which doeurent the proposed applicable ccrrective actions to the discrepancies and pregran 1
deficiencies.
As a result of disagrecrents between mcobers of the ECTG and the HCTTG regarding the proposed TVA corrective actions to reselve the employce cencerns, l
independant experts assessed the issues.
The report docurentino the findincs of consultants relly and Landcrs w3s issued on April 21, 1587
'This report '
partially stated:
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. The current, as-analyzed stress values of TVA Class A smali bcre piping have been reviewed. The nadal points which exceeded 60 percent of either code allowable stress or actual allowable stress were tabulated. There were approximately 2600 nodal points used for the small bore piping analysis of TVA Class A piping. Two and one-half percent of the nodal points had stress ratios which were not capable of meeting the 40-percent reduction on the code allowable stress.
Similarly,1.2 percent of the nodal points had stress atios which were not capable of :reeting the 40-percent reduction on the actual allowable stress.
The report clso partially concluded:
In sumary, the raterial centrol problem is limited to small bore piping. This report deronstrates that tha~ is no technical diUcrence in Class A and Class B piping corrponents, n conclusion, the er.)ineering evaluations demonstrate that the installed small bore pipe and fittings comply with ANSI B31.7c Code requirements when the 40 percent allowable stress reduction factor is used in lieu of NDE.
Thus, plant safety is assured.
This reduction in allowable stress refers to paragraph 1-724 in ANSI B31.7c
- 1971 which states in part:
Unless otherwise required by the Design Specification, and provided all other applicable requirerents of this divisicn (1-274) are ret, the non-destructive exa:rination requirements of this division do not apply to:
1.
Ncn-pressure-retaining raterial:
2.
Seamless pipe end tube, seamless forged socket welding fit :ings, and seamless wrought butt welding fittings 2-inch ncminal pipe size and staller provided that:
a.
The pipe, tube and fittings are made of P num':er 1 or P number 8 materiais that meet all requircrents of one or rore of the standard materials specific nions listed in Tables 1-724 and A-1.
b.
the design stress intensity values (5 ) of Table A-1 used in the design analysis ara multiplied by a fSctor of 0.60.
NOTE: The rajor difference between the smil-bore oipe catcrial recuiremnts of Class A, B and C raterials is the application of non-destructive testing to Cass A traterials.
The three crsviously tentioned Corrective Action Reports (SC-CAR-86-052, S6-055, and C6-C64) docunent the result and ccrrective actions associated with the various discrepancies rated in the three (ECTG, HCTTG, and consultants Velly and Landur) reviews performd at Sequoych.
TVA also performd additional reviews in this area in order to verify tN accuracy of the emplo)ce concerns and to assess the possible effect on the safety of the Sequoyah plant. These revices were M rforN d by Bechtel Structural Integrity As: ociates, and Aptech Engineering.
The highlights of these reviews are st rari:cd balcw.
.6
. Bechtel Auditt The purpose of this audit was:
To verify, by examinaticn of objective evidence, compliance with those aspects of the TVA Quality Assurance Program associated with materials.
Audit to address. program applied both during the construction phase and the operation-phase.
This audit concluded that TVA had generally complied with the connected quality programs and applicable irrpleinenting procedures for material control for both construction and operations.
The exceptions to this compliance were 5 Audit Findings (2 for construction, 3 for operations) and 6 Audit Observations (5 for construction, 1 for operations).
With regard to prograrJatic deficiencies, the Bechtel audit did state:
The findings of this audit do not mveal a deficiency in programatic controls. However, there were instances of implementation errors (i.e.,
incompletely recorded heat numbers, heat numbers recorded on items or documentation partially illegible, etc.) which can create traceability questions recuiring laborious and cost y research and investigation l
efforts.
Structural inQqrity Assecfates (SI A) _ Evaluation.
The three tasks assigned to STA by TVA for its investigation were:
1.
Survey the available docurrentation and industry personnel involved in the construction of other light water reactors built during the same tirra frame as Sequoyah to datertaine the codes and standards invoked for design and construction of those plants and to present the trathods used by other i
utilities fcr traterials control and raintenance of traceability during plant construction.
2.
Obtain a knowledgeable, independent interpretatien of the traccability reouirrT,2nts of the various construction codes alcng with an historical backr/sund of traceability and marking recuirca nts.
3.
By survey of the available data bases, determine whether any cercponent service failure has ever been attributed to improperly dccumented material or resulting f rca a traceability flaw.
This report surcari:ed:
...that materials traceability, although not a code requireN nt, has been important to plant enrers.
Traceability of materials has generally been maintained to a high degree althourh not 100%.
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. Even though a small fraction of material of ouestionable or :ncorrplete pedigree is known to have been installed and placed in service, no failures attributable to such material have been reported.
The methods used by TVA in the design, procurement, and construction of pioing systems for the Sequoyah units appear to have been typical of the day. The heat code traceability questions raised by the Nuclear Safety Review Staff report are nct unique,.Those ouestions relative to Seoucyah do not appear to present an unresolved issue.
hptech RepoQ This report enccmpassed a review of nuclear material manufacturers programs, policies, and practices, as well as nondestructive exaninction versus nuclear classes. This report concluded:
For absolute and unquestionable traceability, the procurement document, the heat code number, and the manufacturer must be known. Also, if any NDE was perform d by someone other than the manufacturer, a separate document was generated showing the NDE method performed and the identity of the material.
The rejection rate of NDE performed on small bore fittings manufactured by forging or machining was less than one percent.
Even today, there are no markings put on small seamless piping products to indicate the class unless the purchasing docucent actually requires this to be done. All manufacturers that were contacted have marked the NDE performed on the raterial since 1980.
Prior to that titre, sorre did and soma did not. We believe that MAVC0 and the material manufacturers preceduras and QA programs root the NAVC0 requirerrents of bc.th ANSI B31.7 and ASME III.
B.
NPC Staf f 9eyiew Sunaury The NRC staf f ccnducted a special team inspection at Sequoyah en June 22-26 ane July 20-24, 1987.
The objective of the inspection was to determirie the accuracy of the inforection contained in the ele: rent report and to deterair.e the adecuacy of TVA's conclusions and corrective actions.
At the conclusion of the inspection effort the NRC staff concluded that in gancral. TVA perfonrad an extensive review of the heat coda traceability issue.
The infomation contained in the element reports was found to accurately scope and review the identified issues.
Fewever, several inadequacies were identified during the HPC team reviews of supporting engineering calculations which were identified as follows:
(1) The review of the supporting pipe calculations identified that TVA has not performed rainimum wall calculations for pipe schedules other than schedule 160.
TVA needs to perform those calculations to ascertain tha' a pressure problem is not present.
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. (2) The acceptance of 2-1/2 percent of nodal points for small-bore piping, based upon the use of actual material properties and thicknesses, is not acceptable.
TVA needs to review those nodal points again and upgrade then, either by performing the additional ilDE, or by adding more supports to reduce the loads, or by replacing the piping.
(3) TVA Design Criteria for Detailed Analysis of Category I Piping Systems, SQN-DC-V-13.3, Rev. 3 provides the loading conditions and stress limits for Category I piping systems in Table 3.1-1.
Footnote 3 of this table states that the allcwable stress levels are given in ANSI B31.1-1967.
TVA's calculations of the allowable stresses for snall-bore piping used ASME Section III, Apperdix I allowables which do nct meet the criteria in SQN-DC-V-13.3.
Since the tine of this itRC inspection, TVA has satisfactorily addressed these three issues by reviewing the piping calculations and upgrading the piping where required.
IV. C0!1CLUSIONS The NRC staff believes that T'iA has adequately addressed the enployee concerns identified in TVA Employee Concern element report l'C-40703, "Heat Code as 1
Related to Material Control." The three issues stated above in the small-bore piping area have been satisfactorily addressed by TVA.
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SECU0YAH NUCLEAR POWER PLANT, UNITS 1 AND 2 SAFETY EVALUATICN REPORT FOR EMPLOYEE CONCERNS ELEMENT RiPORT MC-40705-SQN "QUALITY RECEIVING UNIT" I. Subject Category:
Materials Control (40000)
Subcategory:
Prccedural Centrol (40700)
Elecent:
Quality R:ceiving Unit (a0705)
Employee Concern:
XX-85-027-X02 The basis for Element Report MC-40705-SQN, Revision 1, dated Octcber 31, 1986, is Sequoyah Employee Concern XX-85-027-X02 which states:
"Material inspectors vers not allowed to validate heat numbers of structural steci received ensite as required by procedurc [;] heat No.
7438383 is an example."
This concern was evaluated by TVA cs potentially nuclear safety-related, and only relevant to Sequoyah.
II. Su m rv of Issul The issue definad by TVA is that the concerned individual (Cl) who had been a quality control (QC) inspector felt that during the construction period, there was impedance in the inspection process with regard to heat number validation of structural steel.
The Element Report addresses the impedance issue, but does indicate other areas of concern which resulted from or paralleled this concern (and other concerns by this CI).
A harass. tent issue regarding the CI is being handled by the TVA Inspector General Office by concern number HI-35-005-C01.
Heat number progranTaatic traceability problems are being addressed by concern number MC-40703-SQN.
III. Evaluation, Although seemingly extraneous information appears in the text of the Elemant Report, the thrust of the report is the interviews with QC inspectors by the Employee Concerns Task Group (ECTG).
The NRC staff discussed the details of the report with the ECTG investigators and supervision on January 15, 1987.
Some of the seemingly extraneous information was an attamot to point out oddities in the Employee Response Team (QTC) Re;; ort (XX-85-027-XfA) in the concern area, and with other information pointing cut the rargins between the inpedance concern and the heat number issue of Elecant Report MC-40703-SQN.
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The ECTG investigators interviewed at least ten QC inspectors on possible This interview impedance during performance of heat number validation.
methodology is considered by the NRC staff to be the primary means relevant infor: nation was obtained regarding the concern.
The parametric boundaries of the questions asked by the ECTG of the interviewees should have discerned The ECTG (and the any impedance problems on the part of the QC inspectors.
report) indicated that no inspector had problems validating heat numbers.
As stated by the ECTG, the QC inspectors' only difficulty was with the procedures involved in the validation process which is not centioned in the subject Element Report but was stated by the ECTG to be programmatically addressed in MC-40703-SON.
During the discussion with the ECTG on January 15, 1986. ECTG supervisien indicated that they would probably change Elecent Report MC-40705-SON to point out the procedural problems and the fact that these problems are addressed in MC-40703-SQN.
IV. Conclusion The NRC staff believes that TVA investigation of the concern was adequate and their resolution of the concern as described in Ele, rent Repcet MC-40705, Revision 1, is acceptable. Although the difficult language of the report and side issues identified in the report required some clarification between the NRC and the ECTG ar.d required a working knowledge of the applicable inspection process, the results of the interviews (the ECTG with the TVA QC inspectors) support acceptance of the report.
Any additional clarifica-tion of the report by t..e ECTG should only aid in the readability of the report and not affect its conclusions.
y SEQUOYAH tlUCLEAR POWER PLANT, UNITS 1 AND 2 SAFETY EVALUA110N REPORT FOR EMPLOYEE CONCERNS ELEMENT REPORT OP 30101 "KER0 TEST VALVE LEAKAGE AND CORROSION" 1.
Subject Category:
Operations (30C00)
Subcategory: Mechanical Equiptrent Relinbility/Cesign (30100)
Element:
Kerotest Valve Leakage and Cerrosion (30101)
Errployee Concerns:
I ti-86-285-001 111-85-594-001 XX-85-090 002 XX-85-090-001 EX-85-085-003 This basis for Element Report OP 301.01, dated November 13, 1986, are the following employee concerns:
Concern IN-86-285-001:
Watts Bar Units 1 and 2.
Glcbe valves (kerotest) were received frcn vendcr in a corroded condition due to vendor's hydro of valve and inadecutte drying. These valves leaked after installation. A generic l'CR j
was written to correct this problem, but the full implementation of the NCR disposition is questionable.
Exarples of
-l the systems with these valves are:
Construction Departrrent concern.
CI has no further information, Concern IN-65-594-001:
3/4" kerotest valves (possibly globe); 30 valves inspected with 90* reject-rate; bearings were missing / busted / frozen.
These valves were installt.d throuchout the site
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(Units 1 and 2) and all may not have been identified es evidence of an NCR, hold tags, or further investigations was not kncun.
(Names / details known to QTC).
Concern XX-85-090-C02:
Sequoyah Uni'.s 1 and 2.
Per CI TVA used globe valves (kerotest) extensively in both plarts, Matts Bar and Bellefonte and had leakage and cerrosion problem:.
CI questions the usaga of these valves at Sequoyah - the sister plant - fcr leakage and corrosion problems.
The systems to be checked as examples are CVCS, Safety Injection, RHR and Reac tor Coolant, ett..
01 hhs no further inforiration.
NUC Power Cor.cern."
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2 Concern XX-85-090-001:
Bellefonte all units.
The globe valves (kerotest) need to be checked for corrosion and leakage due to vendors hydro and inadequate drying.
Examples of the systems are:
CI stated that this probicm has existed for six years.
Constuction Department concern.
Cl has no further infornation.
Concern EX-85-085-003:
Kerotest valves are extrenely poor. They seldom seat properly.
Construction Departnent concern.
CI has no additional information.
II.
Sumary of Issue All five of these concerns describe a kerotest valve leakage and corrosion problen in the construction installation process at Watts Bar with generic application to Sequoyah and Bellefonte.
III. Evaluation A problem with kcrotest valves at Watts Der was identified and dccumented in Division cf Cor.struction NCR 2501R dated 10/20/80.
The final report on this problem was issued by the TVA Chief Nuclear Engineer on 4/U/C1.
That report identified several hundred 3/4", 1", and 2" kerotest valves with leakage and corrosion probleas.
The root cause for this leakage and corrosion problem was traced to a factory hydro test of a va h e batch where kerotest allowed the packing to become water saturated and did not remove the wet packing / dry out the valve after the test.
Since the valves were sometines factory tested years before the correncement of operational maintenar.ca, substential opportunity for corrosion of the valve internals was present.
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On the other hand, while seme of the kerotest valves were installed in safety-related systens such as CVC5, sis, RHR, UHI, CCS, and RCS, the operation of these isolation, root, vent, and throttle valves is not required for the scfety shutdcwn of the plant during a loss of coolart accident.
TVA does not consider these valves to perfom a safety function, but believes that the corrosion problen cculd result in a naintenence problem durinc the life of the plant.
As a result, TVA has institrted a maintenance progran to dismantle, inspcct, and replace parts (mainly bearings ard diaphracms) as required for those valves installed at Watts Bar.
The generic application of this problem at Sequoyah was studied as a response to the comitment rade in NCR 2501.
The firal report on this study was issued by the TVA Sequoyah and Watts Par Design Project fianecer on 9/10/81.
That report stated that the kerotest valves' safety function would not be conpromised by problems developing from water saturated stem packing.
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3 Westinghouse provided approximately 1500 kerotest valves to Sequoyah and 1
specified packing replacement after hydrostatic testing.
Another 500 valves were provided by flAVC0 under a spec that did not specify packing repl acement.
It is unknown how many of the NAVC0 contract valves did not have the water sat; rated packing removed before shipment from the factory.
A nechanical maintenance inspection of one kerotest valve in the Sequoyah warehouse in 1986 revealed sone corrosion on the stem and bearing, leading one to conclude that some unknown number of kerotest valves installed at Sequoyah may have some corrosion in the valve internals.
There are two models of kerotest globe valves used at Sequoyah; a "packless" with a diaphragm (packing is used as a backup) and a more typical packing valve.
Both were hydro tested in the factery the same way - with the packing as the external leakpath ty the stem.
A search of all maintenance requests filed at Sequoyah since the plant went into operation, indicated that only one kerotest valve had experienced failure due to leakage or corrosion problems out of appoxinately 2000 kerotest valves installed (.05%).
In addition, the March 1987 search of NPRDS (an INP0 nationwide dcta base for operating r.uclear plants) revealed only 8 kerotest valves had experienced failure due to corrosion out of 191 failures reported over the past 10 years.
Since the NPRDS data base has 3191 kerotest valves listed (out of 110,301 total valves), national failure rate of kerotest valves due to the corrosion was 0.25",.
Based on this statistical data, TVA concluded that the problem Watts Dar hos had with kerottst valves is not generic to Sequoyah, and furtber investigation is unwarranted.
IV.
Conclusicn The NRC staff belieses that the TVA investigation of these concerns ucs adequate and their resolution of these concerns as described in the Elenent Report 30101 is acceptable for restart.
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i SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 & 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERNS ELEMENT REPORT OP 30102 "DIESEL GENERATOR RELIABILITY PROBLEMS" 1.
Subject Category: OPERATIONS (30000)
Subcategory:
MECHANICAL EQUIPMENT RELIABILITY / DESIGN (30100)
Element:
DIESEL GENERATOR RELIABILITY PROBLEMS (30102)
Concern:
XX-85-122-008 XX-SS-122-009 XX-85-122-010 IN-85-323-001 WI-85-100-003 MAS-85-001 The basis for Element Recort OP 30102 - SQN, Rev. 4, dated January 7, 1987, are the following employee concerns.
The following concerns are considered gereric:
IN-85-323-001: "Continuous starting / stopping of Diesel Generators (due to testing) is detrimental to engine parts.
Test progran repuires increased number of tests after a certain number of failures.
CI feels that increased frequency is contrary to vendor recommendations.
CI could net provide specific test numbers.
No additional information available.
NUC Power concern units 1 & 2.
WI-SS-100-003: "Diesel Generators nave reliability problems.
- I stated that correction requires reliability program, a reduction in the number of starts, attention to testing, preventive maintenance upgrading, and more interaction with INPO, other utilities and vencors to establish resolution to problems.
CI has no further information.
Anonymous concern via letter.
The following two concerns apply to otrer plants:
XX-85-122-C09, XX-85-122-010 are Sr:wns Ferry ano Bellefonte c:ncer s that are exactly tne sane as Concern XX-55-122-008 f or Secocyah a c the generic WI-35-100-CO3 Tne following two concerns are Sequoyan specific:
i XX-55-122-0CS: "Diesel Generators ave reliaci':ty prooters.
- I stated that correction requires reliability orcgran, a recuction in the numoer of starts, attention to testing, preventive maintenance i
2 upgrading, and more interaction with INPO, other utilities and vendors to establish resolution to problems.
CI has no further information. Anonymous ;cncern VIA letter.
MAS-85-001: "D/G AC lube oil pump tripped because of possible gasket material in pump."
II.
Summary of Issue Concerns IN-85-323-001 and WI-85-100-003 address the continuous starting and stopping required by test programs for the Emergency Diesel Genera-tors.
The concerns attribute reliability problems with the Diesel j
Generators to excessive starts and stops, a need to upgrade preventive maintenance, and not enough interaction with outside organizations.
Sequoyah specific concern XX-SS-122-008 is almost identical to WI 100-003.
An additional Sequoyah specific concern MAS-85-001 identifies a i
potential problem of gasket material in the Diesel Generator AC lube oil pump.
Concern XX-85-122-009 and XX-85-122-010 are Browns Ferry and Bellefonte concerns that are identical to Sequoyah specific concern XX-85-122-008 and generic concern WI-85-100-003.
i III. Evaluation TVA evaluators reviewed applicable NRC Regulatory Guides, NUREGs, and NRC/TVA correspondence back to the time of licensing.
TVA evaluators identified that numerous NRC documents cating back as far as Generic issue 8-56 in 1977 identified diesel generator reliability as an item of concern.
Evaluators reviewed the recommendations f rom NUREG/CR-0660, "Enhancement of On Site Diesel Generator Reliability," and the comments TVA provided at the request of the NRC. Concerns about no load / light load operations were addressed by separate correspondence between TVA and NRC.
I As recommended, TVA installed air dryers for the air start system anc heavy duty turbocharger dri.a gear assemblies.
Formal training was given to maintenance and other related cersonnel.
Evaluators performed a walkdown of the diesel generator buildings and found housekeeping and oil leaks to be a problem Dust control around diesel generator electrical equipment was identified in N'u REG /C R-C 660 as important.
TVA initiatec corrective action to clean the diesel generator electrical panels and preventive maintenance to maintain overall cleanliness in the diesei generator building.
TVA evaluators revie ed the reliability mi sto y of the Sequoyah Emergency Diesel Generators si ce varch 5, 1930, arc for tre last 100 starts.
Reviews were conducted of EPRI recort ND-4264, ":ailure related to Surveillance Testing of Standby Equipment," and NUREG/CR-4557 whicn cresents an overview of information orovidec by va-icus groups associatec with Diesel Generators at Nuclear Power clants as nail as cc=ents on Generic letter 84-15.
Reviewers notec tnat survefilance testing was l
considered a f actor in diesel gererator failures oy both reports and a j
potential contributor to reducing reliability and lifespan of the diesel l
generator sets.
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TVA evaluators reviewed responses from ASME, EPRI, INPO, and other industry groups which stressed maintenance and analysis 'over, increased -
testing to ' improve reliability.
Vendor responses were reviewed - and-
.particular attention was paid to the. Morrison-Knudsen/GM. response which recommended a reduction in fast starts, use of prelube and prewarm, the elimination of running unloaded,-and improved maintenance.
TVA evaluators concluded that TVA programs included
- 1) vendor.
recommendations for use ' of prelube and prewarm; 2) the elimination of running unloaded; 3) corrective maintenance on cleanliness' in the, diesel generator building and initiation of preventive maintenance 'to maintain cleanliness; and 4) submission of a technical specificati.on change _ (# 107) to reduce the number of starts required by the surveillance instructions.
Concerns IN-85-323-001, WI-85-100-003, and XX-85-122-008 were evaluated as-l being valid with the exception of interaction with outside organizations.
Concern MAS-85-001 was evaluated as not valid because of a lack of supporting findings on lube oil pump failures due to gasket material.
TVA evaluated the root cause as lack of a formal DG trending analysis program and related follow up and lack of adequate continuing housekeeping on the diesel generators.
IV.
Conclusion The NRC staff believes that the TVA investigatico of the concern wa s adequate, and their resolution of the concern as described in Element Report GP 30102 is acceptable for restart.
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SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 ANO 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCC3N? ELEMENT REPORT OP 30107 "GENERAL PAINT CONCERN REACTOR BuiL0iNG" I.
Subject Category:
Operations (30000)
Subcategory:
Mechanical Equipment Reliability / Design (30100)
Element:
General Paint Concern Reactor Building (30107)
Employee Concern:
XX-85-037-On1 The basis for Element Report 30107-SQN, Rev. 2, dated November 18, 1936, is Sequoyah employee concern XX-85-037-001 which states:
Sequoyah Units 1 and 2:
Containment paint coatings (#295 and #305) are not properly maintained.
The integrity of the coatings is being ereded and questionable.
CI is concerned that the paint will curl and pop-up and clog the drains in case of an accident (LOCA) when the temperature and pressure builds up in the reactor. Paint specifica-tions and standards are not followed, especially in reccating of 305.
NUC Power concern.
CI has no further information.
This concern was evaluated by TVA as potentially safety-related.
II.
Summary of Issue The issue defined by TVA concerns the integrity and maintenance of con-tainment paint coatings and the potential safety implications should deficiencies in these areas exist.
TVA also evaluated the adequacy of coatings applicater and inspector cualificatien.
III. Evaluation TVA evaluators reviewed maintenance records, plant procedures, cesign documents, anc recuirements and commitments for the coating system.
General inspections of tne containment coatings were conducted for both units.
Coating applicaters and inspectors were interviewed to ascertain work practices and the extent of their knowledge and training pursuant to i
coating requirements. TVA found that cercrete c: stings were adhering well with scme exceptions wnere significant me:harical camage coupled with water seepage had caused calamination of the ::cc:at.
Inspection of tne steel containment liner #:und areas of tcpccat delamination ard areas I
ahere total fi'm thic(nesses were exceeded.
a 'ack of inscection and testing detail in the mainter,ance instruction for a00iication and repair
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of containment coatings aa s fcund to Oe a c:ntdcuting factor to these deficiencies.
A d::i t i c na l l y, TVA noted tnat n'.y
'imited correctiva maintenance of containment the lack of an upper tier documentcoatings nad been performed in the past due to l
periodic inspection and maintenance. that defines a formalized program of Although deficiencies in containment coatings were identified, TVA concluded frem review of design documents that no impact on safety existed.
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Regarding coating applicator and inspector qualification, TVA concluded that programs in place were adequate.
It was noted, however, that the applicator qualification program lacked procedural control.
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TVA concluded that the issue presented by employee concern XX-85-037-00i regarding maintenance of containment coatings at SON was valid.
TVA has defined an extensive corrective action program, including field repair of defective containment coatings and programmatic changes, to address ider.tified deficiencies.
IV.
Conclusion The NRC staf f believes that the (VA investigation of the concern was adequate, and their resolution of the concern as described in Element Report OP 30107 is acceptable for restart, t
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SEQUOYAH NUCLEAR POWER Pl. ANT, UNITS 1 AND 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERNS i
ELEMENT REPORT OP 30115 "HARDWARE NOT PROPERLY IDENTIFIED" I.
Subject Category:
Operations (30000)
Subcategory:
Mechanical Equipment Reliability and.0csign (30100)
Element:
Hardware Not-Properly Identified (30115)
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Concern:
XX-85-102-005 The basis for Element Report 30115 - SQN, Rev. 1, dated December 12, 1986, is Browns Ferry Employee Concern XX-85-102-005 which states:
Hardware is not properly identified in the field.
A person needs a drawing to identify it.
This concern was evaluated by TVA as r.ot safety related and potentially applicable to Sequoyah (generic).
II.
Summary of Issue TVA perceived the issue to be incorrect or missing equipment identifica-tion tags at Sequoyah.
1 III. Evaluation The TVA evaluator reviewed the Sequoyah Special Maintenance Instruction (SMI) for system walkdowns to determine actions underway to identify problems with equipment tags.
Additionally, an interim report by the Office of Nuclear Power Configuration Management Survey Team and the results of evaluations at the Watts Bar and Browns Ferry sites were reviewed.
Since several hundred missing tags for the systems or portions of systems being walked down have been identified, the concern is con-sidered valid.
Corrective Action Tracking Document (CATO) OP 30115-SQN-01 was written to track the corrective action items belcw:
Complete prior to restart, tagging deficiencies identified and classified as "Restart" by TVA's DB&V Program.
Corrections prior to restart will be limited to components identified by the main control flow and control drawings, and corrected under the DB&V Program.
Initiate a long term corrective action plan to reflect the unique identification of mechanical, electrical, ar.d !&C components on the necessary drawings to allow the components to te reflected in occ-cedures and identified in the field.
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2 The Element Report also states that CATO OP 30115-NPS-01 was written thE corporate configuration manager's of fice to address to the resolution of data base and drawing discrepancies.
IV.
Conclusion The NRC staff believes that TVA's investigation of the concern was adequate, and their resolution of the concern as described in Element Report OP 30115 is acceptable for restart.
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SEQUOYAH NUCLEAR POWER PLANT, UNITS'1 & 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERNS ELEMENT. REPORT OP 30201 "POSSIBLE LACK 0F WATERTIGHT CONDUIT AND t
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Subject i
Category:
Operations (30000) 1 4
Subcategory:
Electrical and Communications (30200)
Element:
Possible lack of Watertight Conduit and Connection (30201)
Concerns:
MAS-85-002 MRS-85-005 1
TAK-85-001
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i The basis for Element Report OP 30201-SQN, Rev. 3 dated January 8,1987, are the following employee concerns:
TAK-85-001:
Guidelines for use of RAYCHEM (coating) on Class 1E work are unclear and instructions not consistent.
MAS-85-002:
Adequacy of RAYCHEM on 2-FCV-43-77, MRS-85-005:
2-FSV-43-77 did not have the proper RAYCHEM application.
II.
Summary of Issue Concerns TAK-85-001, MAS-85-002 and MRS-35-005 identified that splices using RAYCHEM kits may be inadequate at Sequoyah and the procedures at Sequoyah controlling splices may be inadequate.
III. Evaluation The TVA investigations concluded that TAK-55-001 was a valid concern.
Interviews substantiated that Modification and Addition Instruction (M&AI) 7 "Cable Termination, Splicing, and Repairing of Damaged Cable,"
is unclear when trying to specify the correct application of heat shrin(
insulation on a splice.
The TVA investigation concluded that MRS-55-005 and MAS-85-002 were valid concerns.
Subsequent Work Requests (WRs) have corrected the identified problems.
TVA performed visual inspection cf a representative sample of RAYCHEM-installations and found evidence of another deficient application on an envirormentally qualified motor coerator.
The TVA investigation found instances where material requisitten forms were not being attached to work packages as required by Sequoyah proce-dures.
This created a situation were the specific RAYCHEM kit used could not be identified.
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The TVA investigation. identified a potential for similar problems as those identified in NRC Information Notice (IEN) 86-53,. "Improper Installation of Heat Shrinkable Tubing, dated June 26, 1986, existing at Sequoyah. The.
evaluation process of IEN 86-53 needs to continue and required corrective action finalized, The TVA corrective action plan is to identify and correct deficient splices on Environmentally Qualified (EQ) Equipment and Class 1E splices at containment penetrations prior to restart. TVA has provided a justifi-cation for not identifying and correcting splices on the balance of. Class 1E equipment until af ter restart.
This is in accordance with Sequoyah Standard Practice (SQA) 191, Rev. 2, "Evaluation of Operational Readiness Prior to Plant Startup," which indicates that Post restart action has been identified, responsibility assigned, scheduled, and placed in an appropriate management tracking system.
IV.
Conclusion The NRC staff believes that the TVA investigation of the concern was adequate, and their resolution of the concern as described in Element Report OP 30201 is acceptable for restart.
SEQUOYAH NUCLEAR POWER PLANT, UNITS 1 AtlD 2 SAFETY EVALUATION REPORT FOR EMPLOYEE CONCERNS ELEMENT PEPORT OP-30202 "FIVE PERCENT LOW VOLTAGE PROBLEMS" I.
Subject Category: Operations (30000)
Subcategory:
CableandConduit(30200)
Elenent:
Five Percent low Voltage Problems (30202)
Employee Concerns: XX-85-122-004 XX-85-122-005 MAS-86-004 The basis for Element Report OP 30202-SQN, Rev. 3, dated December 8, 1986, are the following employee concerns:
"XX-85-122-004 - Sequoyah - a 5 percent voltage drop at each plant causes problens by cycling diesel generators unnecessarily (which degrades reliability) and causes too many plant shutdowns. TVA compensates by operating buses at higher than normal voltage ratings, anticipating voltage reductions, stressing equipment and components unnecessarily, and reducing component life and reliability.
CI stated that there was inadequate voltage regulations for buses ~.
q Cl has no further information. Anonymous concern via letter."
XX-85-122-005 - Browns Ferry - is worded identical to the above concern but specifically for Browns Ferry and was transmitted as generic to other plants.
MAS-86-004 - Sequoyah - Potential equipment damage as a result of station over voltage.
II.
Summary of Issue This element report evaluates the concern of five percent low voltage starting of the energency diesel generators and of compensating for the five percent low voltage by operating the safety-related buses at higher than normal voltage levels.
The concern of diesel generator starts is alleged to cause unnecessary plant shutdowns and place undue stress and wear on the diesel generators thereby reducing their reliability.
The concern of higher than normal voltage supplied to safety-related equipment is alleged to have overstressed and reduced the life and reliability of plant equipment.
The scope of this report was limited to evaluation of alleged diesel generator starts, plant shutdowns due to five percent low voltage, and the presence of higher than nornal voltages on the safety-related 6.9-kV and 480 volts shutdown boards.
The issue of diesel generator reliability is addressed in element 4
report 30102 "OG Reliability."
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III. Evaluation The TVA evaluator reviewed the K Forms to determine the areas of concern.
Documents were reviewed to determine the requirements, archived copies of weekly Sequoyah Nuclear Plant Surveillance Instruction, SI-3 were reviewed to determine the past voltage conditions, and interviews were held with cognizant individuals in SQN division of power system operations, operations; the operations procedures group, SQN design services (DNE),
plant modifications group, and with a former supervisor of SQN electrical maintenance.
The issue of cycling cgs unnecessarily due to undervoltage conditions was considered not valid because no evidence of such starts could be located in the LER history.
The issue of TVA compensating for 5 percent voltage' drops by operating buses at higher than normal voltage ratings could not be verified.
Ample evidence existed to demonstrate that shutdown boards have been operated at low and high voltages (based on ANSI C84.1 and TVA requirements).
Anticipation could not be demonstrated, but inadequate requirements and procedures allowed the adverse voltage conditions to develop and be sustained in some cases. The concern of high voltage on buses was found to be valid.
The issue of overstressing and degrading the reliability of equipment was found to be valid.
The issue of inadequate bus voltagi' regulation was found to be valid because of the excessive voltage swings on the start buses cnd the shutdown boards when they are connected to either the grid or to the SQN turbine generators.
The issue of potential equiprent damage as a result of station bus overvoltage at SQN was found to be valid because of numerous instances of voltage on both the 6.9-kV and 480 volt shutdown boards in excess of the maximum limit recommended in ANSI C84.1.
These repeated excessive voltages have the potential for degradation of energized electric motors and other equipment through the mechanism of overheating and eventual insulation breakdown.
As a result of this report a Significant Condition Report (SCR) EEB86147 R0 was initiated on November 21, 1986, to address these conditions.
At present, TVA has identified which class IE AC electrical equipment is susceptible to overvoltage and has revised SI-3 to reflect the requirements of ANSI C84.1.
IV.
Conclusion The NRC staff believes that the TVA investigation of the concerns was adequate, and their resolution of the concerns as described in Element Report OP 3020?
is acceptable for restart.
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