ML20150C158
| ML20150C158 | |
| Person / Time | |
|---|---|
| Issue date: | 10/13/1987 |
| From: | Robert Pettis, Potapovs U Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20150C150 | List: |
| References | |
| REF-QA-99900403 NUDOCS 8803180045 | |
| Download: ML20150C158 (33) | |
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE, CALIFORNIA REPORT INSPECTION INSPECTION NO.: 99900403/87-03 DATES: 6/15-18 & 7/27-8/6/87 ON-SITE HOURS: 264 CORRESPONDENCE ADDRESS: General Electric Company Nuclear Energy Business Operations ATTN: Mr. N. L. Felmus, Vice President and General Manager 175 Curtner Avenue San Jose, California 95125 ORGANIZATIONAL CONTACT: Mr. J. J. Fox, Senior Program Manager TELEPHONE NUMBER * (anM Q M -M CA NUCLEAR INDUSTRY ACTIVITY: General Electric Company's Nuclear Energy Business Operations (GE NEBO) is engaged in furnishing engineering services for domestic and foreign nuclear power plants.
A ASSIGNED INSPECTOR: 61 _ 44 , /d '~ f R. I/'i Ptttis, SpeciarPfojects spection Section ate
($PIS)
OTHERINSPECTOR(S): R. P. McIntyre, SPIS S. Alexander, SPIS P ,, Consultant P. Eshleman, Consultant APPROVED BY: 9M -
LO-13-I7 U.Potapovs, Chief,(PIS,VendorInspectionBranch Date l
INSPECTION BASES AND SCOPE:
A. BASES: 10 CFR Part 21 and 10 CFR Part 50.
B. SCOPE: The purpose of this follow-up inspection was to review allegations involving potential deficiencies in design control activities within the Quality Assurance program at GE San Jose, during the period March 1978 l to April 1982. In addition, the status of previous inspection findings l
was also reviewed.
PLANT SITE APPLICABILITY: Potentially multiple plant sites, including River Bend, TVA Units 17-22 (identified by GE as cancelled), Perry 1/2, Nine Mile Point 2, Hope Creek 1/2, Grand Gulf 1/2, Limerick, Clinton, and Susquehanna 1/2.
880318o045 871020 PDR OA999 EMVGENE 99900403 DCD
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS 9AN .10SE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: ) AGE 2 of 27 A. VIOLATIONS:
None.
B. NONCONFORMANCES:
Contrary to GE Engineering Operating Procedure 42-6.00, "Independent Design Verification," dated April 30, 1981, GE did not perform a seismic analysis, as required by the GE Problem Review Board in a letter dated November 17, 1980, extending the original seismic qualification data, performed in 1978, to a 1980 revised design configuration of the reactor mode switch and its interface with the isolation casing (87-03-01).
C. UNRESOLVED ITEMS:
None.
D. STATUS OF PREVIOUS INSPECTION FINDINGS:
- 1. (Closed) Nonconformance (87-01-01) ,
GE issued Engineering Change Notice (ECN) NJ-17436, dated June 2, 1980, without a technical justification to delete the requirement for glyptal coating of GE Electrical Metallic Tubing (EMT) because of unavailability.
GE's response to this item of nonconformance indicated that GE decided to use Purchase Part Drawing 175A9666 to improve the identification and control of various sizes of EMT as part of a company decision to identify commercially available materials on Purchased Part Drawings. GE stated that the statement "BAKED ON CLEAR GLYPTAL" was simply copied from a GE commercial catalog and was not a necessary design requirement for any EMT used in nuclear control and instrumentation panel assemblies manufactured by GE. In May 1980, GE (San Jose) learned that EMT with baked on glyptal was
! no longer commercially available and therefore deleted the statement j from the drawing via ECN NJ17436. As a result, this item is closed.
- 2. (0 pen) Unresolved Item (86-01-07)
GE Engineering Practices and Procedures (EP&P) 5.38 Addendum 4, dated December 1975, required that a tracking system and status log of deferred verifications be maintained. The inspectors l
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA l REPORT INSPECTION NO.: 99900403/87-03 RESULTS: ' AGE 3 of 27 verified during NRC Inspection 86-01 that the first entry was made :
in the status icg for deferred verifications in May 1977. At that time, it could not be detemined whether verifications had been deferred before May 1977 since the status log did not contain entries of any deferred verifiestions prior to that date.
During the 86-01 inspection, GE committed to perform an extensive review of deferred verifications from inception through May 1977 to positively demonstrate closure of deferred design verifications.
During an NRC review of this effort during the June 15-18, 1987, inspection, GE reviewed all 15,322 Engineering keview Memorandums (ERMs) generated from inception to May 1977 to identify ERMs containing a deferred verification statement. As a result, 962 ERMs were identified which affected 3423 design documents. A computer search of these documents, performed on GE's Engineering Information System (EIS), was then used to identify the current deferred verifica-tion status of the affected documents which resulted in only three documents identified as "u" (unverified). At the end of the inspection, GE comitted to researching further the status of these three documents to verify closure. A DBASE III computer program was used by GE to produce a list of deferred verifications based on criteria previously established by the NRC inspector. The criteria established was based on safety-related shippable components produced by GE NEBO, San Jose, for use on domestic nuclear power plants.
This search produced approximatley 130 design documents of which the NRC inspector selected six for further review by GE. GE's review consisted of a manual search of documentation (ERMs, ECNs, etc.)
necessary to demonstrate positive opening and closing of each deferred verification throughout the history of each document. This review also verified the current status as now reported in the EIS.
In a few instances, the six documents selected for review by the NRC related directly to items referenced in Mr. Stokes' summary of Mr. Milam's work record and are associated with Limerick, Susquehanna, and the Shoreham nuclear plants. These six items are:
Item Document No. ERM System / Component 1 283X569 BMA 0743 Reactor Vessel Top Guide 2 851E378 CMA 111 Reactor Protection System Elementary Diagram
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS 9AN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: ) AGE 4 of 27 Item Document No. ERM System / Component 3 828E375TF AMC 0057 RWC and Recirculation Bench Board 4 865E152 AMC 0871 RHR/HPCI I Relay Vertical Board 5 237X574TN AMC 0600 HPCI RLY Vertical Board 6 133D9538 AMC 0568 RCIC RLY Vertical Board During the 87-03 inspection, GE informed the inspectors that it had completed this review. This unresolved item will remain open pending NRC's review of GE's actions during the next inspection.
- 3. (Closed) Unresolved Item (87-01-09)
GE'sProblemReviewBoard(PRB)statedinaletterdatedJune13, 1980, that a modified reactor mode switch should replace those previously shipped and that the design changes required were to be documented via a Field Disposition Instruction (FDI) document. The FDI referenced by the PRB was not available for review during the previous NRC inspection.
GE's response to this unresolved item stated that the correct reactor mode switch is now installed in all boiling water reactor (BWR) plants. This installation was initiated as a replacement in BWRs in late 1983 and 1984, based on a GE product improvement redesign and the issuance of NRC Infortnation Notice 83-42. Because the GE response did not fully address the concerns of unresolved item 86-01-09, this item remains open pending a further review of the 1980 upgrade data and the 1983 product improvement redesign including all affected FDI's.
I l Ins)ection Findings - The inspectors reviewed documentation contained in Potential Reportable Condition (PRC) files 80-57 and 83-22. PRC 80-57 dealt with the 1980 modification and improvement to the housing assembly. After engineering evaluations were performed b GE, which included two meetings of the Problem Review Board (PRB)y ,
recommendations were made to improve the mode switch housing
, assembly. These recommendations (identified as Product Improvements i by GE) were made via Engineering Change Notices, not FDIs, at
! applicable plants. The 1983 redesign of the reactor mode switch l -
l L
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS 9AN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: ,
PAGE 5 of 27 came as a direct result of problems experienced at Susquehanna Unit 1 concerning loss of certain scram signals during switching of the reactor mode switch. Subsequent to Susquehanna's notification of this problem to the NRC, IE Information Notice 83-43 was issued which initiated action to correct these problems. The inspectors verified that a replacement mode switch was sent to all affected utilities via a Field Disposition Instruction (FDI) er a Field Deviation Dispositon Request (FDDR).
Based on this review, this unresolved item is considered closed.
Additional infonnation concerning the reactor mode switch is presented in Section E.6 of this report.
- 4. (0 pen) Stokes Report Section 1.6 Engineering Review Memorandums (ERMs)
"In the first week of November 1978, the following line was part of an entry: Bill Millard said either he would sign the ERMs or I (Sam) could forge his signature to them." (Clarificationaddedby Mr. Stokes.)
Inspection Findings - During the 87-01 inspection, discussions were IIeld with Mr. Millard at which time he denied any such statements concerning "forging" of his signature. Because specific details were not available as to the ERM referenced by Mr. Milam's work record entry in November 1978, the inspector was unable to verify whether Mr. Milam signed his own name or whether Mr. Milam "signed for" Mr. Millard. This item will remain open.
- 5. (0 pen) Stokes Report Section 1.7 Elementary Diagram Drafting Effort t "Continuing with a problem of similar nature on November 14, 1978,
! a letter to C.W. Hart on the subject of the CNV connection has an interesting paragraph. It seems that the CNV elementary diagram drafting effort was subcontracted to an outside firm, the Power Division of C.F. Braun & Company, in Alhambra, California. When l completed, the diagrams were provided to the General Electric System Engineers for signature. The system Engineers felt that they were not being given sufficient time for review and refused to sign the
! documents. The documents were later signed by the C&EE CNV Engineer, without review." N l
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS -
9AN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 6 of 27 Inspection Findings - During the 87-01 inspection discussions were held with Mr. C.W. Hart Mr. Milam's supervisor during this period, who stated he had never received the November 14, 1978 letter. In addition, specific examples of insufficient review times could not be identified from the coments contained in the ERMs or the discussion with Mr. Hart. This item will remain open.
- 6. (0 pen) Stokes Report Section 6.2 Unauthorized Signature Changes "Mr. Milam wrote a letter to W.M. Barrentine on April 14, 1982 about unauthorized, post signature changes. In this letter, Mr. Milam states that R.L. Reghitto made an authorized change to ERM AML-2997 without Mr. Milam's knowledge and in direct conflict with specific instructions."
Inspection Findings - A discussion during the 87-01 inspection with Mr. Barrentine, in the presence of Mr. Barton Smith, GE counsel, inquired as to what actions were taken concerning this subject.
Mr. Barrentine stated he had not received Mr. Milam's letter of April 14, 1982. He also stated that he was not aware of anyone else who might have known about the letter and also might have acted on it in his (Mr. Barrentine's) place while he was on business travel.
This item will remain open.
- 7. (0 pen) Stokes Report Section 6.3 Let_ter to Management "On May 22, 1982, Mr. Milam wrote Mr. Barrentine a letter and included a copy of his work record while working for Mr. C.L. Cobler.
In this letter, Mr. Milam requested Mr. Barrentine to read about the on-going underworld of C&ID and says he tried to comunicate some of these things to Mr. Barrentine on several occasions but was discouraged by Mr. Barrentine's managers and attitude. Mr. Milam says:
Since you no longer hold my form 38 (a standard threat), I have nothing further to fear from either you or your conspiratorial managers. I hope, by sending you this Record, to give you a glimpse into that hidden world of uncontrolled bootleg activity we all know so well, j
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY Bl'SINESS OPERATIONS 9AN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: ' AGE 7 of 27 Mr. Stokes also stated that Mr. Barrentine was the manager of the Nuclear Control & Instrumentation Product Design Operation (NC&ID) of(C&ID). He was Mr. Hart's, Mr. Cobler's, Mr. Reghitto's, Mr. Strambach's, Mr. Koslow's, and Mr. Wortham's supervisor.
Mr. Milam had been notified of his layoff when this last letter was written and his reference to form 38 had to do with the constant threat of layoff if you did not go along with the system. He did not."
Inspection findings - In a discussion during the 87-01 inspection with Mr. Barrentine stated he never received the letter in question nor the portion of Mr. Milam's work record. This item will remain open.
- 8. (0 pen) Stokes Report Section 5.13 River Bend Excluded Equipment List "Mr. Milam's work record included a nonapproved form titled PWA No.
1229LD, Revision IJ for River Bend. This document, which is dated February 5,1982, was caused by an excluded equipment list which was sent to the utility Gulf States Utilit:es Company, by the NRC. The second page of this document states that there is no controlled tracking system for vendor idt-ntification of these devices and that a complete item by item search of the entire River Bend database would be necessary. GE felt that the scope of such a search was prohibitive and furthemore was not considered to be necessary.
Excluded equipment as referred to in this list is equipment which I
has been found at other facilities to be so deficient.that plant safety is seriously in question. GE neither admitted nor denied that this equipment was installed at River Bend."
Inspection Findings - The objectives during this portion of the l
! inspection were to detemine, to the extent possible, the following:
1 (1) What items on the Excluded Equipment List (EEL), which was I
prepared and submitted to GE by Stone and Webster Engineering Corporation,are used by GE NEB 0 in what plants and applications.
(2) Whether the use of those items is appropriate for the particular application in light of the problems or deficiencies with the items that caused them to be listed.
(3) How nomal GE controls identified the problems associated with listed items.
l ORGANIZATION: GENERAL ELECTRIC COMPAN'Y i NUCLEAR ENERGY BUSINESS OPERATIONS I SAN JOSE. CALIFORNIA l
REPORT INSPECTION RESULTS: > AGE 8 of 27 NO.: 99900403/87-03 (4) The effectiveness of GE NEB 0 and product department systems for disseminating information on problems with equipment in preventing inappropriate use.
(5) How and to what extent, if any, GE was negligent in not responding to the request from Stone & Webster to confirm that listed items were not used at the River Bend Plant.
To make these determinations, the inspectors reviewed the following documentation:
(1) A GE computer generated listing of all purchase part drawings (PPDs), called"CATNIP."
(2) ThefileofGEServiceInformationLetters(SILs).
(3) The Index of Field Deviation Disposition Requests (FDDRs),
Field Disposition Instructions (FDIs), Potentially Reportable Conditions (PRCs)(under10CFR21),andtheIndexof"Germane-to-Safety" (GTS) Notifications (Considered by GE to be non-reportable under 10 CFR 21, but still a safety concern to GE customers).
(4) A listing from the GE Service Representative in Philadelphia of Service Advice Letters (SALs) issued by GE Power Systems Management Business Division (PSMBD) for components for which there was a PPD annotated to reflect which were on file with the NEB 0 electrical and instrumentation and control (I&C) engineering department.
Although additional information must be reviewed before conclusive findings can be reached with respect to the stated objectives, the following points were evident from the infonnation reviewed during the inspection:
(1) Review of the CATNIP revealed that there were PPDs for components of the same manufacturer and with the same model number as many of the EEL items. (Note that in some cases, the EEL lists only manufacturer and model and in others, particular lots or manufacturing periods are specified, requiring review of other documentation to determine if the components in question were purchasedand/orused.)
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 9 of 27 (2) A review of the SILs by the inspector indicated that, in many cases, NEB 0 had advised customers of problems associated with EEL items and had provided recommendations for corrective measures.
(3) The NRC inspector noted that although NEB 0 procedures call for issues that are to be the subjects of SILs to be screened for safety ramifications and if any, forwarded to the PRC Coordinator for 10 CFR 21 screening, many SILs were issued in addition to 10 CFR 21 reports on the same issue.
(4) Not all SALs pertaining to EEL components were on file with GE NEB 0 Electrical Engineering although it was learned that NEB 0 was taking actions to be placed on the distribution list for all SALs.
(5) The inspector observed no evidence of a formal system to ensure that SALs received by NEB 0 would be screened by other than QA for applicability to particular plants or generic applicability.
In addition, no evidence of a formal distribution and/or action tracking plan for QA to use for SALs was observed.
Additional documentation was requested and intended to be reviewed along with the above to aid in the objective determination.
Some information was made available and was requested from product departments; however, none of the following was reviewed during the inspection.
(1) Complete listings from the three GE product departments that supplied EEL components, for which there was a PPD, of all SALs issued by them on those components.
(2) A listing (for comparison with the lists of SALs issued) of which SALs are held by GE Electrical Engineering, QA, or elsewhere particularly as relating to EEL components.
(3) The file of customer correspondence held by the GE project engineer for River Bend and Nine Mile Point-2, individual FDDRs selected from the FDDR Index pertaining to EEL components, and the file of individual FDIs selected from the FDI Index pertaining to EEL components.
(4) Individual PRC and GTS evaluations /dispositiens selected from the PRC/GTS Logs and SALs pertaining to EEL components.
,,,y- - - - - - - ,w-- - -- -,-,-- -
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS RAN JOSF. CALIF)RNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: > AGE 10 of 27 (5) A complete "Where Used" printout from the GE Engineering Informa-tion (EIS) data base for PPDs of EEL components.
(6) Procurement documents for components with EEL model numbers to deter 1nine if the lot numbers or particular manufacturing period of concern were included.
(7) QA documents, such as receipt inspection records for components with EEL model numbers to determine if the lot numbers or particular manufacturing periods of concern were included.
As discussed above, the NRC inspector drew some interim conclusions on the basis of the information reviewed during this inspection as discussed previously. However, determining conclusive findings with respect to the stated objectives will require a further review of the documentation listed above. As a result, this item will remain open.
E. OTHER FINDINGS AND OBSERVATIONS:
Background Information As stated previously, NRC Inspection Report Nos. 99900403/86-01 and 87-01 did not attempt to address all of the allegations raised by Mr. Milam and Mr. Stokas, but rather, a representative sample of potentially more significant allegations was selected for review. However, all allega-tions received by the NRC are being addressed and will be documented in future inspection reports. Previously, the area of deferred design verification was addressed which represented the allager's major concerns (as noted during an NRC interview with the alleger in April 1986). As stated in Section D.2 of this report, this item is open and will be reviewed during the next inspection.
This inspection report primarily focuses on the follow-up of items rf nonconformance and unresolved items identified in NRC Inspection Report No. 99900403/87-01. This was accomplished in part by a review of GE's responses to such items (GE's letters to the NRC dated February 5 and March 5,1987) in addition to a formal review of all documentation supportive of GE's response. In addition, several new issues in the areas of fire protection and the use of unverified documents used in preparing GE licensing documents, including the FSAR, were introduced l
during this inspection as a result of Congressman Edward Markey's April 10, 1987 letter to NRC Chairman Lando Zech. It should be noted that issues involving fire protection were not contained in Mr. Stokes'
- report, but have been extracted directly from Mr. Milam's work record.
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS RAN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: ' AGE 11 of 27 The representative sample of allegations inspected is sumarized below, along with the results of the NRC review of each item. The inspection was composed of personnel interviews, examination of applicable files, records and procedures.
- 1. Halon-Based Fire Protection Systems During the inspection, the inspectors reviewed GE's response to Grand Gulf licensing question 013.22, (April 1981) relating to Halon-based fire systems. Halon-based fire system are important safety features in control rooms since they suppress flames without destroying the operator's ability to breathe. One such question posed by the licensee required GE to verify that the Halon system installed at Grand Gulf to protect the Power Generation Control Console (PGCC) floor sections is designed to provide a 30 percent concentration. GE's response stated that the floor section fire suppression system is designed to provide an initial concentration of 6 percent to 7 percent, by volume of Halon within 10 seconds of initiation and sustain a 20 percent concentration for 20 minutes.
GE's response to the licensee raised questions from Congressman Markey as to the minimum concentration standard in addition to the fire systems furnishing only two-thirds coverage and the possible effect this may have on the operators ability to control the plant in the event of a severe fire.
Inspection Findings - The NRC does not have a requirement of 30 percent halon coverage for control rooms. NFPA 12-A, Standard on Halon 1301 Extinguishina Systems,1985 Edition, publishedT the National Fire Protection Association, specifies minimum design concentrations (given as percent by volume) for Halon 1301 ranging from 5.0 percent to 8.2 percent depending on the type of fire (fuel) involved. The 1973 Edition of NFPA 12A specified two larger concentrations - 12 percent for carbon disulfide and 20.0 percent for hydrogen - but the rest of the concentrations given are similar to those listed in the 1985 Edition. The Standard also states: "In addition to the concentration requirements, additional quantities of agent may be required to compensate for any special conditions which would affect the extinguishing efficiency. Carbon disulfide and hydrogen ignite easily and are difficult to extinguish. They have been dropped by specific reference from the standard and are covered by paragraph 2-3.2.2(d), which gives minimum design concentrations required to extinguish normal fires involving several flammable liquids and gases. Design flame extinguishment concentrations not listed shall be obtained by test plus a 20 percent safety factor, and minimum design concentrations shall be 5 percent.
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS RAN JOSE. CALIFORNI A REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PA_GE 12 of 27 The GE PGCC control room is constructed of prefabricated and prewired modular units consisting of control cabinets and consoles mounted on steel floor panels with cables and cable harnesses located in under-the-floor steel cable troughs. The total flooding Halon 1301 fire suppression systems that are the subject of this question are for the protection of these under floor cable troughs. The Halon requirements address concentrations (percent volume of these under-the-floorcabletroughs). There are no nuclear power plants with total flooding Halon 1301 suppression for the entire control room .
volume.
The concept of Defense-in-Depth as applied to nuclear power plant fire protection involves several different levels of activity and concern. The first step in the program is fire prevention. The second step is to provide fire detection and fire suppression capabililty in the event a fire should occur despite fire prevention activities. The third step is to ensure by means of separation and passive fire protection that even if a fire should occur and be promptly extinguished, it will not prevent safe shutdown of the plant because of fire damage to safe shutdown cables and components.
The defense-in-depth theory also affects the second step with respect to minimum design concentration for Halon 1301. After the Browns Ferry fire in March 1975, the NRC staff devoted considerable time to development of a broad spectrum of fire protection guidelines for nuclear power plants. During this time GE was developing and NRC was reviewing and approving, minimum design concentrations for Halon 1301 protection in under-the-floor cable trough for PGCC installations. For some of these installations design concentrations as high as 30 percent were agreed upon. These higher concentrations were stipulated to give a greater margin of safety because of any unknown factors that could not be easily quantified. In addition, it was reasoned, the volumes to be protected and resultant increased installation costs would be small, and the increased hazard to personnel from Halon 1301 leaking out of the cable trough into the control room volume would be negligible. ,
In more recent installations the need for such conservatism in specifying minimum design concentrations three to five times higher than those specified by NFPA 12A has been reevaluated. With installations using improved cable constructions that meet the fire-retardant criteria of IEEE Standard 383, the NRC staff found no technical justification to continue with the earlier conservative approach calling for substantially higher Halon 1301 concentration.
In addition, the staff found that the original rationale was not based t
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS.
RAN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: > AGE 13 of 27 upon arguments that can now be justified as technically compelling.
The NRC staff therefore, concluded that use of the higher concentra-tions of Halon 1301 in the newer plants would not result in a corresponding increase in plant safety.
On this basis, the NRC staff concluded that reducing the concentra-tion of Halon 1301 in the PGCC under-the-f;oor cable troughs to those generally accepted concentrations contained in NFPA 12A would not adversely affect fire protection nor would it affect the operator's ability to control or shutdown the reactor in the event of a severe fire in the control room. The Halon 1301 protection for the PGCC cable trough is only one of the components of the defense-in-depth approach to the provision of fire protection for these control rooms.
It constitutes protection above that nonna11y provided for control rooms and more than satisfies those minimum design concentrations stipulated in NFPA 12A. Therefore, reducing the design concentra-tion from those earlier specified levels, which are now understood to have been unnecessarily conservative, does not constitute any reduction of protection.
Discussions with the GE staff and a review of GE's design files revealed that Halon concentrations were changed because cable material insulation with lower concentrations and longer soak times providing the required fire suppression was in compliance with NFPA 12A recomendations. Included in this review were descriptions of site-specific test procedures to verify the adequacy of the suppression material as applied at each plant installation. No relaxation of the fire protection design standards was noted in this review. As a result, this item is considered closed.
- 2. Kaowool vs. Sand Another area potentially affecting the protection of control room instrumentation in the event of a fire was covered in an internal GE memorandum dated May 23, 1960 with regard to the fire stop design requirements for Grand Gulf 1/2 and Clinton 1. The memorandum stated that a combination of metal barriers and Kaowool, both co'.ered with RTV Rubber would constitute the fire break design in the control room under-the-floor cable troughs. This snemorandum also indicated concern about the inability of Kaowool to fill the cable interstices and that Kaowool may be too easily removed. The memorandum also stated that unless specific NRC approval is obtained, this design approach may be unsatisfactory.
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CAllFORNIA REPORT INSPECTION N0.: 99900403/87-03 RESULTS: PAGE 14 of 27 ,
During early discussions between GE and NRC concerning fire protection for the under floor cable troughs in the PGCC, one design concept that was considered involved filling the cable trough with sand.
After considerable discussion of the idea in 1976-1978, all parties agreed that the potential disadvantage outweighed the benefits of using sand as a fire stop, especially considering the low risk of fire occurrence and associated damage to control room cables.
Therefore, the concept of filling the cable troughs with sand was not adopted by GE and was never required by NRC. GE's current design of the fire stops is based upon test data obtained by GE during tests conducted at the University of California at Berkley.
The design concepts are presented in design concept document NE00-10466A titled "Power Generation Control Complex Design Criteria and Safety Evaluation." This document and amendments are referenced in the FSAR's of the GE plants utilizing PGCC equipment. The criteria for the fire stop material is referenced as 3 inch minimum of a refractory material. The tests at the University of California are included as a reference in this document. The refractory material used in these tests was No. 20 sand. A refractory blanket material is currently utilized as a fire stop material in openings which do not have cables passing thru the opening. However, in the areas where cables are present an RTV foam material is applied as the fire stop material and sealant between the cable trough and is utilized throughout current plants as a fire stop. GE stated the existence of an analysis documenting the acceptability of the RTV foam material in lieu of using No. 20 sand, as utilized in the original test program. However, NED0-10466A does not reference this alterna-tive material. Because GE considers NE00 documents to be licensing l
documents only (not design documents), GE does not intend to revise l NED0-10466A to reflect the substitution of RTV foam material for j No. 20 sand.
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- Where Kaowool had been installed, the design details were reviewed
! and found acceptable by the NRC. Fires that are caused by earth-quakes generally involve rupture of flanmMe liquid and gas storage tanks and piping distribution systems. Since these hazards are not present in nuclear power plant control rooms and since these facilities are designed and constructed to resist and prevent unacceptable damage from earthquakes, earthquake-induced fires are not anticipated and are not included in the design criteria for control rooms.
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN 00SL CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 15 of 27
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The inspectors recomended that NED0-10466A be revised to reflect the RTV foam material as an acceptable alternative to using sand for applicable plant installations. GE stated their position regarding the revision of technical licensing topical reports is: (1)asa rule, technical topical reports are not revised after they have been approved by the NRC, (2) the topical reports are not intended to provide design requirements to any design groups within GE, and (3) the topical reports provide an acceptable method for addressing a generic issue or a way of meeting an NRC staff requirement based on best available information at the time. GE stated that these licensing topicals can be referenced in specific plant FSARs as a preapproved licensing document.
The utilization of the RTV foam material, used at Grand Gulf and Clinton 1 (and possibly other plants), remains an open ites and will be reviewed further during the next inspection.
- 3. Unverified Documents Stokes Report Section 3.9 "Mr. Milam had a disagreement with George Stramback on 3-3-81 over the use of unverified documents to verify FSARs. Mr. iiilam felt that they should note the use of unverified documents to verify FSARs and formally notify the responsible engineer that his unverified documents were used to support an FSAR quality review.
Mr. Stramback felt that the verification status of the documents that were used was not important. Mr. Reghitto was in agreement with Mr. Stramback."
l Stokes Coment: Unverified documents should not be used to verify I any other design documents. This includes the FSAR, which is used l
by the NRC Comissioners to grant the license. 10 CFR 50 Appendix B, I
Section I, "Or are those of (ganization," says a) assuring that an that the quality appropriate assurance quality assurance functions program is established and effectively executed and (b) verifying, such as by checking, auditing, and inspecting activities affecting safety-related functions, to ensure that those activities have been correctly performed.
Inspection findings - The NRC inspector verified that part of Mr. Milam's job at the time was to perfont quality reviews of the l FSAR sections assigred to him during which he was to identify the l
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS f SAN JOSL CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 16 of 27 design documents used to support the descriptive text and' document that correlation, during which time Mr. Milam identified design documents which may not have been in a verified status at the time of his review. GE stated that they were not aware of any FSAR updating caused as a result of an unverified document change.
The FSAR text is a description of the plant as depicted in other GE, Architect-Engineer, and Utility documents. It provides a narrative explanation which is reviewed by the NRC in addition to the other documents which are either required to be supplied to the NRC or are available upon request. It represents a functional level explanation of the drawings, symbols, and other documents which are the basis for the design and demonstration of plant adequacy and is the responsi-bility of the utility to maintain. If the utility maintains the total responsibility, GE sends updated documents to keep the FSAR current. These updates may be used for clearing of design verifica-tion or other reasons. It should be noted that the FSAR is a constantly changing document up to plant licensing and undergoes many amendments. The GE design document process (in compliance with the Quality Requirements of 10 CFR 50 Appendix B) controls the genera-tion of documents for first time use and any necessary later changing of infonnation contained in those documents. If a document was issued for use without verification being complete (deferred verification), it is scheduled and tracked until the verification process was completed. In most cases the document when finally verified, required no change from the already issued document.
However, if a change is necessary, as a result of completion of verification after the document was first issued, the Change Control process outlined in GE E0P 55-2.00 would be initiated.
This E0P process authorizes and initiates the modification to the FSAR document, as well as, the design document itself. The Engineering Change Authorization (ECA) process and its related documents include a Change Impact Check Sheet which requires identification in the Change Impact Effects section of FSAR type changes. Once the need for the FSAR change is identified, the affected FSARs are put on the ECA and scheduled for completion. The task for completion is then scheduled and tracked in GE's Work Planning and Schedule System (WPSS).
In an attempt to verify this system the NRC inspectors selected eight ECAs for review, generated by GE for Grand Gulf, during the 1981 period as covered by the work record with a total of four ECAs noted as potentially affecting the Grand Gulf FSAR (ECAs 801203-1A,
1 ORGANIZATION: GENERAL ELECTRIC COMPANY l NUCLEAR ENERGY BUSINESS OPERATIONS uN JOSF. CALIFORNI A l REPORT INSPECTION .
NO.: 99900403/87-03 RESULTS: PAGE 17 of 27 810609-1A,810430-2A,and810615-1). ECA C10615-1 issued August 19, 1981 was selected by the inspectors for review in order to verify the documentation path necessary to Cadse the FSAR to be updated.
This particular ECA was generated as part of the Boiling Waw r Reactor Owners Group response to the NRC action plan developed for Three Mile Island as documented in NUREG-0660 and NUREG-0737. The change in question described a logic modification proposed by GE which would have improved the High Pressure Core Spray system automatic response after manual termination of the system.
The inspectors verified that the package initiating such proposed change was complete ar.d contained the required documentation as described in GE E03 55-2.00, issued March 31, 1977. The Change Impact Check Sheet, dated June 15, 1981, was reviewed and contained.
a checkmark next to the column "Licensing SAR's or Topical Report Changes Required" which indicated the potential for such document to be affected by the proposed change. As a result, GE Safety and Licensing Departrrent proposed to the utility (Mississippi Power &
Light and Middle South Energy, Inc.) a change to the Updated FSAR (UFSAR),Section 7.3.1.1.1.3.4. "Logic and Sequencing." This change proposed adding the following section which stated, "The HPCS system can be reset if reactor water level has been restored even if the high drywell pressure condition persists. The HPCS pump can then be stopped and the injection valves closed. Automatic restart I
will occur if the low water level condition returns." A review of Volume 13 of the UFSAR by the inspectors was found to incorporate l GE's proposed change and demonstrated satisfactory compliance to GE l E0P 55-2.00. The UFSAR is the reference document for purposes of I communicating with the NRC such as reporting of deviations from conditions stated in the FSAR and for 10 CFR 50.59 evaluations. The original FSAR, as amended, is considered to be the licensing basis for the plant.
As a result of this review, the inspectors were unable to verify the concerns expressed by Mr. Milam. It was noted, however, that the NED0 documents are used as a reference for FSAR design descriptions, but these same documents are not considered n a design reference l
for the plant designs as developed by the GE responsible engineers.
It was also noted from the discussions held with the GE staff that NED0 type documents are not controlled under 10 CFR 50 Appendix B design control and project application is difficult to established since they are not tracked by project. The NEDO documents are reviewed and approved by management, however the plant designs are 1
1 1
ORGANIZATION: GENERAL ELECTRIC COMPANY
. NUCLEAR ENERGY BUSINESS OPERATIONS UN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 18 of 27 based upon design specifications and referenced to a NEDO document.
From this discussion it was apparent that confusion as to the purpose and use of these documents in the FSAR reviews (where particular designs are presented for NRC approval) may have existed during Mr. Milam's employment while in the Technical Licensing Unit of the Control and Instrumentation Department. As a result of the above, the allegation was not substantiated and is considered closed.
- 4. FSAR Verification Stokes Report Section 3.10 - During FW8112 (the third week of March),
Mr. Milam discovered that they routinely use NED0-10466-A to support FSAR's. He feels that since the purpose of the FSAR review is to show that the FSAR in question is supported by the design, the use of NED0 documents is not appropriate. The NEDO documents do not provide formal support because they are not controlling design documents and therefore do not fonnally define the design. Further items that bothered Mr. Milam were:
- 1. NED0's are not issued through the fonnal engineering review cycle.
- 2. NED0's are not covered by change control.
- 3. Because of 1 and 2 above, neb 0's do not satisfy document requirements for design control and change control of 10 CFR 50 Appendix B.
- 4. Except for very rare exceptions, NED0's do not appear in the GE Engineering Information System (EIS). Thus, the revision status is difficult to establish.
- 5. NED0's do not appear on the project Master Parts List (MPL's).
Thus, project application is difficult to establish.
Coninent: The least the NED0's should do is reflect the design.
Inspection Findings - GE stated that NED0 documents describing designs and systems conform to a proceduralized issue and change control system, as outlined in NE00-22000, and are epproved by GE management, and as in the case of NED0-10466-A, have been accepted by the NRC as a Licensing Topical Report (LTR). The FSAR reviews are instituted to ensure that the FSAR properly reflects plant design. In appropriate cases, NEDO documents are used as an l
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 19 of 27 approved source. The reviewing engineer assures the factual accuracy of the FSAR statement, and determines the appropriateness of using a NEDO document as a source. NEDO accuracy has been appropriately maintained by controlled document revisions initiated by the responsible author and approved by management. NED0 documents describing designs and systems are based on documents produced, issued and revised under the controlled GE QA system (E0Ps). The NEDO documents undergo extensive review with acceptance and signoff by management prior to issue. It is preferred to use design documents in the review of FSARs when an NRC accepted LTR does not exist and when changes to design documents have not yet been reflected in an NRC accepted LTR, as was indicated by Mr. Reghitto's note (attached to Mr. Milam's concern) which was circulated to the responsible GE organization on March 23, 1981.
The following represent NRC inspection findings to Mr. Milam's issues raised under items 1-5 above.
Issue 1 NED0s are not issued through the formal engineering review cycle.
Inspection Findings NE00s are issued and revised through a fonnal GE document control system. They are not processed through the "formal engineering review cycle" because they are not formal design documents which conform to 10 CFR 50. GE design requirements and NRC commitments which may appear in NED0s are incorporated by the responsible design engineer into GE design documents which are controlled by the design document control system (E0Ps).
Issue 2 NED0s are not covered by change control.
Inspection Findings As discussed previously in Stokes Item 3.9 (Section E.3 of this report) and in response to issue 1 above, GE does not require NEDO documents to be included in a formal design change system.
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNI A REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PA_GE 20 of 27 Issue 3 Because of issues 1 and 2 above, NEDO documents do not satisfy docu-ment requirements for design control and change control of 10 CFR 50 Appendix B.
Inspection Findings Mr. Milam's statement was correct. GE classifies NEDO type documents as licensing related only and are not used for design purposes.
As discussed in the response to Issue 1 above, the design requirements requiring design and change control under 10 CFR 50 Appendix B are included in design documents that are issued and revised in full compliance with criterion III of Appendix B.
Issue 4 Except for very rare instances, NED0 documents do not appear in the EIS. Thus, the revision status is difficult to establish.
Inspection Findings Since NED0s are not formal design documents, they need not appear in EIS. The revision status of NED0s is available through either the NEDO library or the engineer responsible for a given NEDO. Issue 5 NEDO docun. ants do not appear on project MPLs. Thus, it is difficult to establish project application.
Inspection Findin;s Since NED0 documents are not formal design documents, inclusion in the project MPL is not necessary or appropriate. If project applica-tion is of interest, each FSAR includes a listing of all NED0s referenced.
With regard to Items 4 and 5 above, it should be noted that the FSAR is included in both the MPL and EIS and therefore NE00s referenced in the FSAR are tied to both systems. Considering this explanation, Mr. Milam's concerns about the appropriate use of NEDO documents was not substantiated, and this issue is considered closed.
ORGANIZATION': GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS RAN JOSE. CALIFORNIA REPORT INSPECTION N0.: 99900403/87-03 RESULTS: PAGE 21 of 27
- 5. Subcontractor Perfomance Stokes Report Section 4.27 - On October 26, 1981, Mr. Milam received work packages for B21 FCD update for Grand Gulf, Perry and River Bend. Two Grand Gulf device lists were unverified and not so noted on OMTEC Documentation.
Inspection findings - GE stated that part of Mr. Milam's job when interfacing with subcontractor personnel was to review inputs received from the subcontractor to confim their accuracy and acceptability.
It appears Mr. Milam, in identifying certain inaccuracies in the subcontractors input, was properly performing his assigned work task. In order to verify that no unverified documents exist today for Grand Gulf, Perry and River Bend, the NRC inspectors reviewed several B21 series Functional Control Diagrams (FCDs) to verify their completion status. During the period in question, the following FCDs and associated Device Lists (DLs) were reviewed.
Project FCD # DL #
River Bend B21-3090 82ET45AA B21-3050 828E443AA B21-3060 851E225AA l Perry No Information available
) Grand Gulf B21-3090 828E445BA l B21-3060 828E444BA A review of all affected Deferred Verification Status Change Notices (DVSCNs) associated with the opening and closing of each deferred verification was performed by the inspectors and observed to be complete. One example reviewed (ECN NJ14507, dated September 12, 1980, for Document DL 828E444BA), prepared by OMTEC (Formerly Jet Consultants, Ir.c.) demonstrated independent review and approval by GE staff personnel including the notation "Deferred Design Verification" in the verification, statement block of the ECN. A further review indicated OMTEC as being an approved vendor and so l
! noted on the GE Approved Vendor List. It was also verified that l OMTEC had a GE approved QA program as early 1979, with regular GE audits conducted as part of GE's Vendor Surveillance Program. The NRC inspectors verified that all deferred verifications reviewed for l
this ECN were subsequently cleared prior to plant start up as l
required by GE procedures. As a result, this allegation was not i substantiated and is considered closed.
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ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS RAN JOSF. CALIFORNIA )
REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 22 of 27 !
- 6. Reactor Mode Switch During this inspection Potentially Reportable Condition (PRC) files 80-57 and 83-22, as well as related concerns involving the Rundel-Gould reactor mode switch were reviewed. PRC 80-57 included a ;
June 13, 1980 letter documenting the concerns of the Problem Review Board (PRB)andsolutionstoidentifiedproblemsaswellasa November 17, 1980 letter which documented the follow-up meeting of the PRB and its coments concerning the reactor mode switch. The alleger, Mr. Sam Milam, was a member of both PRB meetings. PRC 80-57 dealt primarily with potential deficiencies caused during the panel mounting of the reactor mode switch which has previously been discussed as part of NRC Inspection Report Nos. 99900403/86-01,and 87-01.
PRC 80-57 The concerns raised in PRC 80-57 were dispositioned by GE as not affecting the safe operation of the switch, but did result in product improvements for panel mounting. The most significant concern involved disassembly and subsequent reassembly of the reactor mode switch neck for attachment to the switch isolation housing. This operation had the potential for positioning the switch out of phase. However, because this could only be caused by a deliberate act, it would be easily recognized prior to and during testing following panel assembly and field installation. This potential problem would still not disable the function of the switch or actual reactor scrams. GE modified.the mode s.4 itch to isolation housing attachments to prevent assembly errors and by doing so I
affected the seismic response characteristic of the switch. There-i fore, at a minimum, a documented analysis should have been performed l
to extend the original seismic testing (performed in 1980) to the t current design as earlier stipulated in the November 17, 1980 letter documenting the follow-up meeting of the PRB. Since GE could not produce the seismic analysis during the inspection, the inspectors concluded that the switches provided by GE did not have full seismic
) qualification covering the modifications made. During the inspection, I the NRC inspectors brought this to the attention of GE, and subsequently GE provided the inspectors a Memo of Record dated August 4,1987 which addressed the seismic capability of the changes made to the mode switch. This memo concluded that the reactor mode switch assembly in the control room bench board is seismically qualified as documented in the original seismic qualification l
1 e
l ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS _
9AN .10%F. CALIFORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 23 of 27 reports for the switch assembly and the prototype bench board. As a result, Nonconformance (87-03-01) was identified during this part of the inspection for GE's failure to perfonn this analysis, per GE E0P 42-6.00, as reconnended in 1980 by the GE Problem Review Board.
The most significant inconsistencies noted by the inspectors concerned a December 5, 1980 letter from Hendrix/ Grim /Reigel (GE) to W. Barrentine (GE) which stated that, "none of the problems are of a safety concern and our evaluation is that this is not a reportable condition." It was also noted that the letter did not reference the November 17, 1980, PRB letter and also did not address any of the recomendations made during that meeting. A May 27, 1981, letter from G. Sherwood (GE) to D. Ferguson (GE) requested documentation for the seismic analysis which was required per the November 17, 1980 PRB letter. In a letter dated July 20, 1981 W. Barrentine and D. Ferguson answered the Sherwood request by stating that the December 5,1980 letter documents the engineering conclusion of the mode switch concerns that no safety hazard exists and that existing reactor mode switch assemblies do not have to be replaced. In addition. an additional seismic test data is required. This staterM , in direct contradiction to the November 17, 1980 PRB lettr m .o sever was addressed or referred to in the December 5, 1980 ,: w .
The final disposition of PRC 80-57 was made in an October 16, 1981 letter written by G. Sherwood of GE Safety and Licensing which concluded that the condition described did not constitute a reportable condition to the NRC under the provn ions of 10 CFR Part 21. The product modifications and improvements as specified by the PRB on November 17, 1980 were implemented per ECNs NJ 21792 and NJ 21793, both dated December 15, 1980, for Susquehanna 1/2 Hope Creek 1/2, Nine Mile Point 2, Perry 1/2, Grand Gulf 1/2, Clinton 1, Liebstadt, Kuosheng 1/2, and Cofrentes.
The chronology of documentation contained in PRC 80-57, as reviewed by the inspector, is as follows:
- 1. May 28, 1980 - memorandumfromD. Taylor (GE) ands.Milam(GE) identifying potential reactor mode switch problems.
- 2. May 29, 1980 - memorandum to W. Barrentine (GE Product Design Engineering) from Quality Systems elevating concerns identified in Taylor /Milam letter.
1 ORGANIZATION: GENERAL ELECTRIC COMPANY
. NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSE. CALIFORNIA ;
REPORT INSPECTION NO.: 99900403/87-03 -
RESULTS: ) AGE 24 of 27
- 3. June 2, 1980 - memorandum from W. Barrentine/D. Ferguson (QA NC &
ID Manager) to G. Sherwood, Safety and Licensing requesting their evaluation of reactor mode switch problems.
- 4. June 13, 1980 - PRB meets to discuss problems and develop solutions for the reactor mode switch. Additionally, GE stated that replacement of previously shipped switches should be considered.
- 5. November 17, 1980 - PRB meets again to discuss reactor mode switch modifications and additional actions including recommenda-tion for scismic analysis to determine effects of change. No recomendation for replacement of previously shipped switches.
- 6. December 5, 1980 - memorandum from Hendrix/ Grim /Reigel (GE Engineering) to W. Barrentine documenting PRB evaluation input.
Memo states that none of the problems from the June 10, 1980 PRB are of a safety concern and all of the proposed solutions are being completed except #1.
- 7. May 27, 1981 - memorandum from G. Sherwood to D. Ferguscn requesting infonnation on mode switch replacements and documenta-tion of the seismic analysis by engineering.
- 8. July 20, 1981 - memorandum to G. Sherwood from D. Ferguson stating that no change-outs are needed and neither is a seismic analysis because the deficiencies and changes were insignificant.
- 9. September 4, 1981 - memorandum to G. Sherwood stating that potential problems with the reactor mode switch do not affect the safety functions and Product Design Engineering's opinion 1s that this condition is not reportable under 10 CFR 21.
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- 10. October 16, 1981 - memorandum from G. Sherwood closes out PRC 80-57 stating that the condition is not reportable under 10 CFR 21.
PRC 83-22 PRC 83-22 addressed the reportability and significance of the reactor mode switch design and manufacturing deficiencies identified at Susquehanna Unit 1 before initial startup. While engaging the reactor mode switch, certain scram signals were disabled in positions in which the sc ams were required to be operable. As a
ORGANIZATION: GENERAL ELECTRIC COMPANY ,
NUCLEAR ENERGY BUSINESS OPERATIONS
%AN JOSF. CALIFORNIA REPORT INSPECTION N0.: 99900403/67-03 RESULTS: PAGE 25 of 27 result, GE initiated action immediately to correct the deficiencies with notification made of the problem to the NRC by Susquehanna.
This action initiated NRC IE Information Notice 83-42.
The NRC inspectors verified the issuance of Field Disposition Instructions (FDI) and Field Deviation Disposition Request (FDDR) notifying affected licensee's to replace the original mode switch with the new design. The plants notified were River Bend 1, Clinton 1, Susquehanna 1/2, Grand Gulf 1, Perry 1, and Hope Creek 1.
Additional Questions Concerning the Reactor Mode Switch Included in Congressman Markey's April 10, 1987 letter to NRC Chairman Lando Zech, were questions concerning aspects of the GE reactor mode switch. These questions are presented along with the NRC findings.
l Duestion 1 Did the redesign of the mode switch include seismic testing and actual field installation testing?
Inspection Findings - The mode switch mounting arrangement modifica-tions and product improvements did not include any new documented seismic testing outside of what was performed for the original design. Field testing verifying continuity and contact pickup for each mode was performed by both GE and the licensees.
Question 2 Were the original causes of the misalignment deficiencies identified?
What were they?
Inspection Findings - The alleged misalignments identified in PRC-80-57 involved the disassembly and improper reassembly of the i
early Rundel-Gould switch neck for attachment to the isolation l housing. The inspectors observed that any incorrect disassembly I would be irrrnediately identified during initial panel assembly and testing and therefore any mode switch misalignments would be imediately detected.
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4 ORGANIZATI0n: TNERA'. ELECT'lIC COMPANY neC'.Ev ENUGY BUSDE% OPLMTIONS
_ SANJOSE.CAq0RNIA REPORY NSPECTION NO.: 99900403/87-03 l RiSULTS: PAGE 26 of 27 1
Question 3 What reinstallation procedures were prepared to assure that the initial misalignment problems do not occur when the switches are replaced? How frequently are reactor mode switches replaced?
Inspection Findings - Any misalignment would be imediately determined in the GE checkout and the licensee preoperational tests prior to fuel load. The misalignment is not a credible problem for any near term or currently operating BWR plant because these plants contain the 1983 installed mode switch with the improved mounting arrangements which prevent this from occurring. All mode switches in-place today are expected to operate for the 40 year plant life.
Question 4 Were plant owners who may have to engage in corrective action on-site made aware that during replacement of mode switches problem could occur with the switch contacts that cause them again to be misaligned unless specific corrective action procedures were followed? If yes, please provide documentation of which plant owners were identified and when. If no, why not?
Inspection Findings - For current operating plants this is not an occurrence that is likely to happen based on the response to Item 3 above.
Question 5 Was the mounting for each new switch installed by GE properly qualified for compatibility with the new switch? What is the basis for this conclusion?
Inspection Findinas - The actual panel mounting of the mode switch has not changed. The mounting improvements concern only the switch connection to the isolation can assembly which is perfomed by GE prior to shipment to the site.
Question 6 What other isolation systems exist as backup to the mode switch?
ORGANIZATION: GENERAL ELECTRIC COMPANY NUCLEAR ENERGY BUSINESS OPERATIONS SAN JOSF. CALITORNIA REPORT INSPECTION NO.: 99900403/87-03 RESULTS: PAGE 27 of 27 Inspection Findings - The inspectors interpreted "isolation systems" to mean "scram signals." Typically the following scram signals are not affected by the mode switch operation, and therefore can be considered as a backup. In addition, these signals would not become disabled should problems develop during operation of the switch,
- a. Turbine stop valve closure
- b. Control valve fast closure
- c. Scram discharge volume high level
- d. Containment high pressure
- e. Vessel high pressure
- f. Vessel low level
- g. Main steam line high radiation Question 7 Pending completion of necessary corrective action on the mode switches, what additional tests or inspections has the NRC conducted to assure the backups will provide accurate infonnation to operators?
Inspection Findings - No corrective action on the reactor mode switch is required by the NRC. The current surveillances, annunciators, and instrumentation available to plant operators would l detect any reactor mode switen malfunctions.
I As a result of this review of the reactor mode switch concerns, this item is considered closed.
- 7. Subcontractor Signature Concerns Congressman Edward Markey indicated concerns over possible
! improprieties of permitting subcontractor personnel to sign GE l drawings produced by the subcontractor. During the inspection the
- inspectors reviewed several ECNs and connection diagrams produced by i OMTEC (formerly JET Consultants, Inc.), a GE subcontractor during l the 1978 period, to determine if OMTEC personnel had signed for GE l employees. ECNs NJ04925, NJ04926, and NJO9992 were examined in I
addition to connection diagrams 807E480AD, and 807E371AB. These documents were noted as being originated by an OMTEC employee, and in all cases the review and approvals were performed by GE staff personnel.
As a result of this review, no example could be found to support the Mr. Markey's concerns. This item is considered closed.
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gf: IDS Code IE:09 VIB Reading DRIS Reading JPartlow EMerschoff UPotapovs RPettis /
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NAME :RPettis:tt* :UPotapovs* :EMerschoff* : G.p... :JPart : :
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DATE :10/05/87 :10/07/87 :10/13/87 :10//3/87 Pl' : 10Ao /87 :
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