ML20206F861
| ML20206F861 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 04/10/1987 |
| From: | Berry K CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 8704140382 | |
| Download: ML20206F861 (35) | |
Text
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Consumers Power Kenneth W 8erry MMM Director yggg g Nuclear licensing General Offees: 1945 West Parnalt Road Jackson. MI 49201 e (517) 788-1636 April 10, 1987 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT l STATUS OF RESOLUTION OF FINDINGS FROM INSPECTION REPORT 86-13 1 Nuclear Regulatory Commission IE Inspection Report 86-13 dated November 4, 1986 described 12 findings related to Big Rock Points EEQ program implementa-tion. Of these 12 findings, 5 were classified as potential enforcement /
unresolved items and 7 were classified as open items. This letter provides an update on the resolution of each of the 12 findings. It also responds in detail to NRC letter dated March 12, 1987 which required Consumers Power Company to submit a detailed program plan (including time schedules) describing how we plan to resolve the NRC staff's cable qualification concerns. Our efforts to expedite resolution of the NRC Staff's cable qualification concerns prior to the 1988 refueling outage will result in significant changes to our engineering staff priorities. These changes in priorities I will require delays in other Integrated Plan project schedules. An addendum to the February, 1987 Integrated Plan Update will be submitted about mid-May reflecting the schedule changes. l
, UNRESOLVED ITEM 50-155/86-13-01 Inadequate implementation of SER/TER commitment regarding the qualification of polyethylene and butyl-rubber insulated cables.
RESPONSE
Our present cable qualification is based on NRR comments on previous qualification methods for Big Rock Point cables. Previous qualification methods have included elevated temperature tests on cable specimens removed from the plant in 1975, analytical evaluations based on material properties, identification of the types and physical condition of installed cables, and reliance on testing of similar cables (generic testing).
8704140382 870410 PDR ADOCK 05000155 G PDR O
.- Nuc1:ar Regulatory Commisnion 2 Big Rock Point Plant Status of Resolution of Findings From IEIR 86-13 April 10, 1987 In 1975, as part of the Big Rock LOCA Task Force, samples of fourteen various types of cable were removed and tested at elevated temperatures. The lowest temperature at which a sample failed was 135"C (300*F). Some of the samples passed at temperatures as high as 165*C (354*F). Big Rock Point's peak accident temperature is 113*C (260*F). The test records refer to some of the cable codes covered in the existing qualification files. Other cable types were also tested, but available records do not, in all cases, reference cable codes. This demonstrates that a significant cross section of Big Rock cables are capable of withstanding temperatures far in excess of conditions which would be experienced at Big Rock Point. The Big Rock LOCA Task Force also performed analytical evaluations to substantiate that the materials used for Big Rock cable insulation were capable of withstanding radiation doses from both normal conditions and accident conditions. A complete report of the LOCA Task Force effort was submitted to the NRC on May 2, 1975. NRC letter dated May 30, 1975 stated the report provided reasonable assurance that the equip-ment will operate as required in the postulated LOCA environment.
In 1980, Consumers Power Company contracted with the plant's original Architect / Engineer to investigate the condition of cables in the Big Rock Point Plant. This investigation report also refers to some of the cable codes covered in the existing qualification files. The report of this investigation offered the following conclusions:
" Practically every type of 1961 cable ... plus more recently installed cables were observed in several preselected areas including the ventila-ting, containment, penetratior, switchgear and compressor rooms....
The ' condition' of all cables was excellent. In fact, about the only way that any difference between the 1961 and the more recently installed cables could be detected was on the basis that the 1961 cables had no visible surface marking compared to the surface printing on the newer cables.
Based upon ' sight and touch', this observer did not find any cable deterioration or defects....
In retrospect, this installation and its condition and environment represents an almost ideal cable storage situation ..."
This investigation demonstrates not only one expert's opinion that Big Rock cables were in good condition; it also demonstrates Consumers Power Company's attempt to identify age related degradation of cables so corrective action could be taken if necessary.
In the Franklin TER dated February 18, 1983, these qualification efforts were not accepted. Consumers Power Company submitted Revision 3 to Big Rock Point's EEQ Program dated March 15, 1982 which committed to test certain M10487-0002A-NLO2
.. Nucisar Regulatory Commission 3 Big Rock Foint Plant Status of Resolution of Findings From 1EIR 86-13 April 10, 1987 categories of cable. On April 26, 1983, the NRC provided an SER along with the TER and requested our plans for qualification or replacement of equipment.
Consumers Power Company's May 31, 1983 letter stated: "A comprehensive review program is being conducted for items in NRC Categories I.B and II.A and will be scheduled as part of our integrated assessment program which will be described in our upcoming June 1, 1983 submittal." The June 1, 1983 submittal described a comprehensive integrated assessment of all open issues. Details of the EEQ plan was provided in this submittal. The EEQ plan concluded by stating it: " meets the intent of 10CFR50.49 requirements and replaces all previous EEQ commitments."
In June 1984, an NRC audit of EEQ files stated the electrical cables: "should be tested as originally planned, or a more positive effort to show qualifica-tion through testing already completed should be initiated."
In response Consumers Power Company provided an EEQ program update on October 31, 1984 with a table which listed TER item numbers and Consumers Power Company proposed resolution. The resolution for three cable types was to show qualification through testing which had already been completed. One cable type was replaced. Consumers Power Company did not test and does not plan to test cable samples from the plant. Consumers Power Company believes that our cable qualification represented a positive effort as required by the June 1984 NRC audit report.
In September, 1986, the NRC performed an inspection of our EEQ Program implementation. The report for that inspection was issued on November 1, 1986. It stated the region would review our past cable qualification efforts with NRR. Based on the NRC's September 1986 inspection report, the outstanding issue regarding cable qualification is the need to establish similarity between cables described in qualification test reports and those actually installed at Big Rock. The report cited four EEQ files which were inadequate.
These four files cover two types of cable insulation: butyl-rubber with a PVC jacket and polyethylene with a PVC jacket.
In order to strengthen the similarity evaluation, original plant construction records were searched. Copies of several field purchase orders, manufacturers certificates of compliance, and material test reports were found. These records provide a complete description of each cable. The purchase order numbers, cable descriptions and cable code numbers all coincide with those in the subject EEQ files. The cable descriptions and specifications are consistent for all purchase orders regardless of manufacturer. On many purchase orders and manufacturers documents, IPCEA S-19-81 is referenced as a specification requirement.
The qualification reports in the subject EEQ files provide detailed descriptions of tested cables. The descriptions in the qualification reports closely match those in our purchase records. EEQ file 2.45 describes a test on a butyl-based insulation per IPCEA S-19-81. EEQ file 2.55 describes a test MIO487-0002A-NLO2
.. ' Nuclear Regulatory Commission 4 Big Rock Point Plant Status oi Resolution of Findings From IEIR 86-13 April 10, 1987 on two conductor 14 AWG stranded copper wire insulated with 20 mils of polyethylene and 10 mils of PVC covered overall with a 45 mil thick jacket of PVC. In both cases, the materials, insulation thickness and jacket thickness are similar to the descriptions provided in plant records. In addition to these records, our plant circuit schedule identifies the cable code for each cable in the plant. This links each installed cable to a specific EEQ file, but not to a specific manufacturer.
In meetings between Consumers Power Company and the NRC staff on March 3 and 4, 1987, the staff stated that since the polyethylene cables in the test report were manufactured by a different manufacturer, similarity of the tested cables to those installed at Big Rock Point had not been demonstrated. The NRC staff stated the formulation (which includes material, manufacturing process, and fabrication) of the cables must also be shown to be similar, and that Big Rock Point's EEQ files do not contain sufficient information to establish such a similarity link. The NRC staff did not state what was necessary to establish an adequate similarity evaluation and indicated that additional deficiencies in the butyl-rubber qualification file exist which were not described in the inspection report.
The NRC staff told Consumers that three options were available that would resolve the staff's cable qualification concerns. These are:
- 1. Perform qualification tests on each of the five vendor / cable types in question to assure acceptability,
- 2. Replace all of the cable in question, or
- 3. Collect additional test or qualification data to show cable similarity.
During the March 4 meeting, Consumers Power Company made two commitments related to resolution of the cable qualification concerns. The first commitment was to submit Justifications for Continued Operation (JCOs) for specific EEQ items which rely on these cables. This commitment was satisfied by our March 6, 1987 letter. The second commitment required Consumers Power Company to choose one of the cptions and submit a detailed program plan within 30 days of startup from the 1987 refueling outage. On March 31, 1987, we met with the NRC to provide the results of our study of options available to resolve this issue.
Under the plan described on March 31, Consumers Power Company will utilize a consultant to evaluate the formulation of each cable insulation type and the l variations between the known manufacturers of Big Rock's cable for outside containment cables. The evaluation is expected to demonstrate that materials, manufacturing processes, and fabrication of the cables are similar and will not affect the cables functional capabilities. This information will be used to establish the link for similarity to cables which were qualified for environmental conditions which envelope Big Rock conditions. Attachment 1 to MIO487-0002A-NLO2
.. Nuclear Rsgulet-my Commission 5 Big Rock Point Plant Status of Resolution of Findings From IEIR 86-13 April 10, 1987 this letter lists every cable serving an EEQ item which utilizes the subject cables. For each cable, the associated EEQ item, the EEQ item location, cable type, length and comments are provided. Attachment 2 provides the post-DBA environmental conditions for the five hostile areas which house EEQ items at Big Rock. The areas outside containment are significantly less hostile than inside containment. Consumers Power Company believes these environments can be classified as semi-hostile as described in NEC letter dated March 17, 1987 since the temperature does not exceed 180*F, radiation levels are much lower and other parameters are normal in most cases. We anticipate similarity will be fully acceptable for cables in these areas, therefore, no cable will be replaced outside containment. The NRC staff concurred that this approach was reasonable and may be an acceptable method of qualification in the March 31, 1987 meeting.
It must be recognized that Consumers Power Company has established a conservative EEQ Program for Big Rock Point. Components have been included on our EEQ list that are not considered to be necessary by the industry nor required by 10 CFR50.49. We, therefore, are removing the Control Rod Drive Scram Solenoids from the EEQ list. The EEQ file for these items is based on FMEA and provides the necessary analysis to remove them from the EEQ list. This action will reduce the remaining quantity of cable inside containment substantially. Additional justification for removing these solenoids is provided in Attachment 3.
The cables which remain will be replaced with new qualified cables. These cables are indicated by an asterisk (*) in the left margin of Attachment 1.
Approximate lengths of cable in the Reactor Building is included in the comments column. The cable replacement effort will include installing new raceway if installation and cable pulling can be done prior to shutdown. This will reduce the scope of outage work and reduce occupational exposure. New raceway will be installed in accordance with original plant specifications.
In cases where radiation fields prevent access during power operation, and in locations where congestion complicates installation of new raceways, we plan to utilize the existing raceway.
During the March 31 meeting with the NRC staff, Consumers Power Company proposed to complete the similarity evaluations by June 30, 1987. Cable replacements were proposed to be completed during the 1988 refueling outage.
The NRC staff indicated that this schedule was inadequate. The staff believes this issue should be resolved aggressively and in a relatively short period of time. Consumers Power Company therefore, is planning an outage to begin by May 29, 1967 (which will last approximately two weeks). During this outage, all lengths of cable containing polyethylene or butyl-rubber insulation located in the Reactor Building will be replaced. Qualification for the lengths of cables outside the Reactor Building will be demonstrated by similarity prior to startup from this outage.
MIO487-0002A-NLO2
.. Nuclssr Rsgulatory Commission 6 Big Rock Point Plant l
Status of Resolution of Findings From IEIR 86-13 April 10, 1987 Big Rock Point's engineering staff is currently reviewing each of the cables to determine if new raceway installation will be required, establishing the l detailed routing of replacement cable, determining the availability of l necessary cable configuration, and writing installation instructions. We expect to have adequate cable and splices in stock at Big Rock Point, from Palisades or Midland, or ava11able from vendors on short notice, but certain l
cable configurations may have to be substituted (eg two separate cables in place of one). Completion of the above engineering detail and related project I documentation and approval to allow prefabrication and new raceway installation is estimated to take no less than four weeks. Upon completion of these activities, it is expected we can begin new raceway and cable installation in the accessible areas the first week of May, 1987. These activities are expected to take approximately four weeks and should include all activities that can be accomplished without the plant in a shutdown condition. The remaining cable replacements splices and terminations would begin concurrent with the plant outage. Engineering and design work will continue throughout the entire period until installation and testing is complete.
Attachment 4 to this letter provides the JCOs for all EEQ items which rely on the subject cables. These JCOs will remain in effect until either qualifica-tion by similarity is established to the NRC staff's satisfaction or the subject cable is replaced with fully qualified cable.
OPEN ITEM 50-155/86-13-02 1
i ltems removed from the EEQ List still relied on in the Emergency Procedures. 1
RESPONSE
Consumers Power Company is reevaluating our commitment to add cautionary statements to Emergency Procedures for items removed from the EEQ list. These procedures refer to items which were never EEQ listed. In the future, more
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non-EEQ items will be referenced in the Symptom oriented E0P's required by NUREG-0737.
Caution statements for the items which once were on the EEQ list could result in the operators concluding all items not having a caution statement are EEQ. This could lead to operator error. Consumers Power '
Company believes a better approach is to indicate in the procedures the items which are EEQ listed. This will be completed in conjunction with emergency procedure rewrite efforts which are currently scheduled for implementation following the 1988 refueling outage per our Living Schedule.
OPEN ITEM 50-155/86-13-03 l
l Use of inaccurate dates in the PACS master file of maintenance and l surveillance activities.
RESPONSE
J From time to time, new activities are added to the PACS listing. Until the activity is actually completed the first time, no actual date can be entered.
The PACS is the activity listing is updated to reflect actual completion dates shortly after accomplished. it is our intention ta continue this practice.
M10487-0002A-NLO2 L
.. uclear R:gulatory Commission 7 Big Rock Point Plant Status of Resolution of Findings From IEIR 86-13 April 10, 1987 OPEN ITEM 50-155/86-13-04 Inadequate program to train key personnel for EEQ activities.
RESPONSE
Big Rock Point's Training Department has written a lesson plan for 16C technicians, maintenance repairmen, and their supervisors. Initial training for these positions was completed on March 23, 1987.
Big Rock Point's EEQ project engineer will attend a vendor training l program covering maintenance and qualification of safety equipment during 1987.
}
At this time, no additional training is planned.
UNRESOLVED ITEM 50-155/86-13-05 l
Limitorque Actuator not qualified to D0R Guidelines for operation in a r design basis accident due to unqualified materials.
R_ESPONSE:
Big Rock Point has replaced V0P-7068 in accordance with Plant Modification SC 86-019. The new operator, a Limitorque SMB-00, was originally installed in our cancelled Midland Nuclear Plant Unit 1 and is fully qualified for inside containment conditions per Limitorque report 600456. No further corrective action is required for this item.
UNRESOLVED ITEM 50-155/86-13-06 Rotork actuator MO-7072 not qualified to D0R Guidelines for operation in a DBA due to unqualified materials.
j
RESPONSE
j This valve actuator was removed from the EEQ list. Our submittal dated Septen.be r 26, 1986 provided justification for this action. No further i
corrective action is required for this item.
1 UNRESOLVED ITEM 50-155/86-13-07 Flow transmitter FT-2162 not qualified for intended function past the November 30, 1985 EQ deadline.
RESPONSE
Since our September 26, 1986 submittal, Consumers Power Company became aware of and obtained a Rosemount report (#D8400323) which qualified series 1153 and 1154 Rosemount transmitters for submergence. This report concludes that if one of these transmitters was submerged in two feet of a chemical solution or water for two weeks with an atmospheric overpressure up to 50 psig at a temperature of 130*F or less, there would be no effect on the transmitters' l performance. At Big Rock Point, FT-2162 (a Rosemount Series 1153) will becomo M10487-0002A-NLO2 l
. _ _ _ - . .- =-- . - . _
- Nuclear Rtgulatory Commission 8 Big Rock Point Plant Status of Resolution of Findings From IEIR 86-13 April 10, 1987 submerged in up to 12 feet of water with an atmospheric overpressure of 27 psig. These test submergence conditions exceed the Big Rock requirements.
The test conditions of 50 psig and two feet submergence imposes a total pressure head of 50.86 psig on the transmitter. The Big Rock conditions of 27 psig and 12 feet submergence imposes a total pressure head of 32.2 psig on the transmitter. In both cases, the transmitter is fully submerged.
BRP EEQ File 2.120 has been revised to state FT-2162 is fully qualified for its intended function.
OPEN ITEM 50-155/86-13-08 1
Deficiencies in the EQ files for various power and control cables relative to IR characteristics.
, RESPONSE:
Consumers Power Company will include performance criteria relative to cable IR characteristics in EEQ files 2.10, 2.16, 2.25; update EEQ file 2.16 to include i the latest test reports from Rockbestos and resolve a discrepancy in file 2.25 concerning qualification time period. Because of the ongoing effort to replace cable inside containment, completion of this open item is forecast to be September 1, 1987.
OPEN ITEM 50-155/86-13-09 Qualification for 30-day operability after a DBA.
I
RESPONSE
The test conditions in EEQ File 3.50 was noted not to envelope the plant profile. An analysis was presented and accepted during the audit justifying a lower peak temperature. It was also noted the test conditions qualified the tape for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Consumers Power Company will revise EEQ File 3.50 to demonstrate 30-day qualification of 3M tape splices. Because of the ongoing effort to replace cable inside containment, completion of this open item is forecast to be September 1, 1987.
OPEN ITEM 50-155/86-13-10 Missing sections in EQ file for states terminal blocks. ;
RESPONSE
f Missing sections have been added to the EEQ file. The subject EEQ file is complete and no further action is required.
ITEM 50-155/86-13-11 Lack of licensee response to concerns in IE Notice 83-72 for Limitorque l actuators. l M10487-0002A-NLO2 I
.. ' Nuclear Rtgulatory Commission 9 i
Big Rock Point Plant Status of Resolution of Findings From IEIR 86-13
- April 10, 1987 i
i RESPONSE:
I f During the September 1986 audit, Consumers Power Company committed to walkdown
); all (6) Limitorque actuators. These walkdowns were completed during our 1987 refueling outage. The NRC Resident inspector was present for some of these f walkdowns. Copies of the completed inspection checklists have been given to the Resident Inspector, do additional action is planned.
t UNRESOLVED ITEM 50-155/86-13-12 Limitorque actuator MO-7080 unqualified to D0R Guidelines due to broken l
terminal block barriers.
'i
RESPONSE
j As stated in our September 26, 1986 submittal, the terminal strip was replaced
! and internal wires were rerouted to eliminate tight bend radius situations.
No further corrective action is required.
i f
l Kenneth W Berry Director, Nuclear Licensing l
i CC: Administrator, Region III, NRC i NRC Resident Inspector, Big Rock Point Plant l Attachments J
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I M10487-0002A-NLO2
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ATTACHMENT 1 Consumers Power Company Big Rock Point Plant Docket 50-155 EEQ CIRCUITS USING POLYETHYLENE OR BUTYL-RUBBER CABLE April 10, 1987 l
l 8 Pages M10387-1406A-NLO2
Page 1 of 8 ATTACliMENT 1 EEQ CIRCUITS USINC PE OR BR CABLE Length Cable No. Scheme EEy Item Location Cable Type (ft) Comment
- CO2-201/8- 601 POS 6626 Reactor Building 3C14 PE 102 Length in Reactor Building = 57' Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
CO2-ZO2/6 601 POS 6628 Pipe Tunnel 3C14 PE 118 Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
CO2-204/5 601 POS 6629 Pipe Tunnel 3C14 PE 112 Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
-CO2-Z06/4 601 POS 6649 Pipe Tunnel 3C14 PE 110 Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
- CO2-Z14/2 601 POS 6653 Reactor Building 3C14 PE 125 Length in Reactor Building = 80' Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
- CO2-Z15/3 601 POS 6654 Reactor Building 3C14 PE 177 Length in Reactor Building = 132' Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
- CO2-Z16/2 601 POS 6655 Reactor Building 3C14 PE 174 Length in Reactor Building = 129' Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
- CO2-Z17/3 601 POS 6618 Reactor Building 3C14 PE 161 Length in Reactor Building = 116' Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
CO2-Z19/1 601 POS 6660 Pipe Tunnel 3C14 PE 113 Operability time for EEQ item = 30 days Containment Isolation Valve Position Indication.
C01-ZO1/16 5414 Spare Outside Penetration Rm IC12 BR 102 Wire is spare in control scheme (26) 1 i
M10387-1406A-NLO2
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Page 2 of 8 ATTACliMENT 1 EEQ CIRCUITS USING PE OR BR CABLE Length Cable No. Schema EEQ Item Location. Cable Type (ft) Comment CO2-Zol/83 5414 Spare Outside Penetration Rm 1C12 PE 88 Wire is spare in control scheme (21a)
- N01-R01/33 B 5414 MO-7068 Reactor Building 3C12 BR 100 All in Reactor Building Operability time for EEQ item = 1 Day Back-up Enclosure Spray Valve
- N01-R01/34 C 5414 HO-7068 Reactor Building 4Cl2 BR 100 All in Reactor Building Operability time for,EEQ item = 1 Day ,
Back-up Enclosure Spray Valve Z01-202/31 5414 Spare Outside. Penetration km_ 2C14 PE 5 Wire is spare in control scheme P12-R01/9 5415 Core Spray Core Spray Pump Ra 3C1/0 BR 222 Operability time for EEQ item = 30 Days Pump Pump / Motor for Recirculation Mode P22-R01/11 5416 Core Spray Core Spray Pump Rm 3C1/0 BR 2 34 Operability time for EEQ item = 30 Days Pump / Motor for Recirculation Mode Pump CO2-N01/1 5418 MO-7066 Core Spray Pump Rm 7C14 PE 255 Operability time for EEQ item = 1 Hour Local Cont. Cooling Water Supply Valve to Post Incident lleat Exchanger.
N01-R01/8 5418 HD-7066 LS Core Spray Pump Rm SCl4 PE 31 Operability time for EEQ item = 1 llour Cooling Water Supply Valve to Post Incident lleat Exchanger.
N01-R01/9 5418 MO-7066 Core Spray Pump Rm 3C12 BR 31 Operability time for EEQ item = 1 Ilour Cooling Water Supply Valve to Post Incident licat Exchanger.
P22-N01/3 5418 MO-7066 Core Spray Pump Rm 3Cl2 BR 215 Operability time for EEQ item = 1 llour Local Cont. Cooling Water Supply Valve to Post Incident Ileat Exchanger.
M10387-1406A-NLO2
Page 3 of 8 EEQ CIRCUITS USING PE OR BR CABl.E Length Cable No. Scheme EEQ Item Location Cable Type (ft) Comment D01-Z01/2 A 5601 For RE09A Outside Penetration Rm IC12 BR 100 Cable inside Reactor Building is not BR Operability time for EEQ item = 1 Ilour Primary Core Spray Actuation on Low Reactor Water Level / Pressure D01-ZO2/1 A 5601 For RE09C Outside Penetration Rm 2C12 BR 100 Cable inside Reactor Building is not BR Operability time for EEQ ltem = 1 liour Primary Core Spray Actuation on Low Reactor Water Level / Pressure
- Z01-202/5 5601 LS/RE09A Reactor Building ICl2 BR 5 All in Reactor Building LS/RE09C Operability time for EEQ item = 1 Hour Primary Core Spray Actuation on Low Reactor Water Level / Pressure
- Z01-ZO3/3 5601 LS/RE09A Reactor Building 1C12 BR All in Reactor Building PS/IC11A Operability time for EEQ item = 1 llour Primary Core Spray Actuation on Low Reactor Water Level / Pressure
- ZO2-ZD4/1 5601 LS/RE09C Reactor Building 2C12 BR All in Reactor Building PS/IC11C Operability time for EEQ item = 1 llour Primary Core Spray Actuation on Low Reactor Water Level / Pressure
- 203-Z04/2 5601 PS/1CllA Reactor Building ICl2 BR 5 All in Reactor Building PS/IC11C Operability time for EEQ item = 1 llour Primary Core Spray Actuation on Low Reactor Water Level / Pressure D01-ZO1/1 A 5602 For RE09B Outside Penetration km ICl2 BR 100 Cable inside Reactor Building is not BR Operability time for EEQ item = 1 llour Primary Core Spray Actuation on Low Reactor Water Level / Pressure M10387-1406A-NLO2
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Page 4 of 8 ATTACllMENT 1 EEQ CIRCUITS USING PE OR BR CABLE
, length Cable No. Ctement
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Scheme EEQ ltem Location Cable Type (ft) 1)01-Zo2/2 A 5602 For RE09D Outside Penetration km ICl2 BR 100 Cable inside Reactor Building is not BR Operability time for EEQ item = 1 Ilour Primary Core Spray Actuation on law Reactor Water Level / Pressure
- 201-ZO2/6 5602 LS/RE09B Reactor Building IC12 BR 5 All in Reactor Building LS/RE09D Operability time for EEQ item = 1 Iluur Priaury Core Spray Actuation on Low Reactor Water Level / Pressure
- 201-203/4 5602 LS/RE09B Reactor Building IC12 BR All in Reactor Building PS/IC11B Operability time for EEQ item = 1 llour ,
Primary Core Spray Actuation on low Reactor Water level / Pressure
- ZO2-ZO4/1 5602 LS/RE09D Reactor Building 2Cl2 BR All in Reactor Building PS/IC11D Operability time for EEQ item = 1 Ilour Primary Core Spray Actuation on Low Reactor Water level / Pressure
- 203-ZO4/3 5602 PS/ICllB Reactor Building ICl2 BR S All in Reactor Building PS/ICllD Operability time for EEQ item = 1 llour Primary Core Spray Actuation on Low Reactor Water level / Pressure
- C01-C15/1 6505 SV-4879 Reactor Building 2C14 PE 155 Length in Reactor Building = 85' SV-4876 Operability time for EEQ item = 1 Hour (FMEA)
Containment Isolation Valve Solenoid Valve
- CO2-YO2/2 6505 SV-4922 Reactor Building 2C14 PE 145 Length in Reactor Building = 60' Operability time for EEQ item = 1 iluur (FMEA)
Containment Isolation Valve Solenoid Valve CO2-ZO3/9 6505 SV-4897 Pipe Tunnel 2C14 PE 146 Operability time for EEQ item = 1 llour (FMEA)
Containment Isolation Valve Solenoid Valve CO2-ZO4/6 6505 SV-4922 Pipe Tunnel 2C14 PE 177 Operability time for EEQ item = 1 liour (FMFA)
Containment Isolation Valve Solenoid Valve M1038 7-1406A-NLO2 1
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5 Page 5 of 8 ATTACHMENT 1 i EEQ CIRCUITS USINC PE OR BR CABLE Length Cable No. Scheme EEQ Item Location Cable Type (ft) Comment
- C15-YO2/2 6505 SV-4879 Reactor Building ICl4 PE 127 All in Reactor Building 3
Operability time for EEQ item = 1 Hour (FMEA)
Containment Isolation Valve Solenoid Valve
- C15-Z01/1 6505 SV-4876 Reactor Building 2C14 PE 72 All in Reactor Building
? Operability time for EEQ item = 1 Hour (FMEA)
. Containment Isolation Valve Solenoid Valve
- C15-ZO2/1 6505 SV-4876 Reactor Building 2C14 PE 151 All in Reactor Building Operability time for EEQ item = 1 llour (FMEA)
Containment Isolation Valve Solenoid Valve
- CO2-C30/1 6514 LS/RE06A Reactor Building SC14 PE 121 Length in Reactor Building = 62' i
LS/RE20A Operability time for EEQ item = 1 Hour Reactor Protection System Trip on Low Steam Drum Level.
- CO2-C30/7 6514 LS/kE06A Reactor Building SC14 PE 121 Length in Reactor Building = 62'
,! LS/RE20A Operability time for EEQ item = 1 Hour j Reactor Protection System Trip on Low Steam Drum Level.
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- CO2-C30/8 6514 LS/RE06A Reactor Building SCl4 PE 121~ Length in Reactor Building = 62' LS/kE20A Operability time f or EEQ item = 1 Hour Reactor Protection System Trip on Low Steam
- Drum Level.
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Operability time for EEQ item = 1 Hour Reactor Protection System Trip on MSIV closure.
- CO2-201/69 6514 LS/RE09A Reactor Building 2C14 PE 183 Length in Reactor Building = 124' Operability time for EEQ item = 1 Hour Reactor Protection System Trip on Low Reactor Water Level, i
MIO387-1406A-NLO2 i
a Page 6 of 8
, ATIACHMENT 1 1
EEQ CIRCUITS USING PE OR BR CABLE Length Cable No. Scheme EEQ Item Location Cable Type (ft) Comment
- CO2-ZO2/18 6514 LS/RE09C Reactor building 2C14 PE 183 Length in Reactor Building = 124' Operability time for EEQ item = 1 Hour Reactor Protection System Trip on Low Reactor Water Level.
002-Z10/6 6514 PS-664 Outside Penetration Rm 2Cl4 PE 86 Operability time for EEQ item = 1 Hour Reactor Psotection System Trip on liigh Containment Pressure.
CO2-Zll/6 6514 PS-665 Outside Penetration km 2C14 PE 80 Operability time for EEQ item = 1 Hour Reactor Protection System Trip on High i Containment Pressure.
- CO2-C30/3 6515 LS/RE06 B Reactor Building SCl4 PE 135 Length in Reactor Building = 62' LS/RE20 B Operability time for EEQ item = 1 Hour Reactor Protection System Trip on Low Steam Drum Level
- CO2-C30/9 6515 LS/RE06 B Reactor Building SCl4 PE llb Length in Reactor Building = 62' LS/RE20 B Operability time for EEQ item = 1 Hour Reactor Protection System Trip on Low Steam Drum Level a
- CO2-C30/10 6515 LS/RE06 B Reactor Building 2C14 PE 135 Length in Reactor Building = 62'
! LS/RE20 B Operability time f or EEQ item = 1 Hour Reactor Protection System Trip on Low Steam Drum Level
- CO2-R01/7 6515 H0/7050 LS Reactor Building SC14 PE 134 Length in Reactor Building = 61' Operability time for EEQ item = 1 Hour Reactor Protection System Trip on MSlv Closure.
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- CO2-ZO1/70 6515 LS/RE09B Reactor Building 2C14 PE 197 Length in Reactor Building = 124'
.I operability time for EEQ item = 1 Hour Reactor Protection System Trip on Low Reactor Water Level.
7 M10387-1406A-NLO2
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Page 7 of 8 ATTACHMENT 1 6
EEQ CIRCulTS USING PE OR BR CABLE a
Length Cable No. Scheme EEQ Item location Cable Type (ft) Comment ,
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- CO2-ZO2/19 6515 LS/kE09D Reactor Building 2C14 PE 197 Length in Reactor Building = 124' l
! Operability time for EEQ iter = 1 Hour
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Reactor Protection System Trip on Im Reactor Water Level.
1 CO2-ZIO/7 6515 PS-666 Outside Penetration km 2C14 PE 100 Operability time for EEQ item = 1 Hour Reactor Protection System Trip on High Containment Pressure.
.; l CO2-Z11/8 6515 PS-667 Outside Penetration Rm 2C14 PE 100 Operability tisus for.EEQ iter = 1 Hour j Reactor Protection System Trip on High i
, Containment Pressure.
- C01-ZO1/9 7501 SV-4892 Reactor Building 2C14 PE 111 Length in Reactor Building = 70' i Operability time for EEQ item = 1 Hour (FMEA)
Containment Isolation Valve Solenoid l
C01-ZO2/11 7501 SV-4895 Pipe Tunnel 2Cl4 PE 127 Operabi!!ty time for EEQ item = 1 Hour (FNEA)
Containment Isolation Valve Solenoid j
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- C01-ZO3/6 7501 SV-4869 Reactor Building -
2Cl4 PE 181 Length in Reactor Building = 130' Operability time for EEQ item = 1 Hour (FMEA) ;
i containment Isolation Valve Solenoid i i
C01-ZO4/7 7501 SV-4896 . Pipe Turunel 2C14 PE 121 Operability time for EEQ item = 1 Hour (tNEA)
Containment Isolation Valve Solenoid 1 .
- C01-205/4 7501 SV-4891 keactor Building 2C14 PE 166 Length in Reactor building = 100' Operability time for EEQ item = 1 flour (FMEA) >
Containment Isolation Valve Solenoid 1
2C14 PE 209 operability time for EEQ item = 30 Days !
001-C26/1 8501 SV-9151 Air Shed Containment Isolation Valve Solenoid
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- M10387-1406A-NLO2 i
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. . _ _ _ _ . _ _ _ . _ _ _ _ _ _-___m. . _ . . - _ _ . _ _ . . . . ._ . = . . _ _ _ . - - . _ . . _ _ ~ _ _ _ _ _ . _ - _ . . _ _ . . . _ -__ m- _ . - . . . , _ -
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Page 8 of 8 2 ATTAQ9 TENT 1 !
a
, EEQ CIRCUlTS USING PE OR BR CABLE C01-C26/2 8501 SV-9151 Air Shed 2C14 PE 209 operability time for EEQ ites = 30 Days Containment Isolation Valve Solenoid l CD2-Zol/3 8501 POS 9101 Air Shed 2C14 PE 220 Operability time for EEQ item = 30 Days j Containment Isolation Valve Position Indication C02-Z03/7 8501 POS 9103 Air Shed 2C14 PE 224 Operability time for EEQ item = 30 Days i Containment Isolation Valve Position Indication i
C26-201/1 8501 POS 9101 Air Shed ICl4 PE 31 Operability time for EEQ item = 30 Days I j Containment Isolation Valve Position Indication C26-ZO3/1 8501 POS 9103 Air Shed ICl4 PE 35 Operability time for EEQ item = 30 Days '
Containment Isolation Valve Position Indication i
I CO2-C26/1 8512 SV-9153 Air Shed 1C14 PE 210 Operability time for EEQ item = 30 Days 1
Containment Isolation Valve Solenoid I :
t CO2-ZO3/22 8512 POS 9102 Air Shed 2C14 PE 221 Operability time for EEQ item = 30 Days Containment Isolation Valve Position Indication CO2-ZO4/4 8512 POS 9104 Air Shed 2C14 PE 225 Operability time for EEQ item = 30 Days ,
J Contaisument Isolation Valve Position Indication 4
f i
M10387-1406A-NLO2
ATTACHMENT 2 Consumers Power Company Big Rock Point Plant Docket 50-155 POST DBA ENVIRONMENTAL CONDITIONS FOR HOSTILE AREAS April 10, 1987 1
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MIO387-1406B-NLO2 1
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1 ATIACHMENT 2 Post DBA Environmental Conditions for liostile Areas f Core Spray Ventilation Parameter Reactor Building Pipe Tunnel Penetration Room Equipment Room Air Shed Peak Temperature (*F) 260 118 140 169 125 Peak Pressure (psig) 27 Atmos Atmos Atmos Atmos 4
I Relative Humidity (%) 100 100 Atmos Atmos Atmos Chemical hone None None None None 6 5 N 4 Radiation (I.) 30 day TID 4.5 x 10 1.87 x 10" 2 x 10 4.29 x 10 2.5 x 10 i
NOIE: Radiation dose represent maximums. Individual components within any area may receive lower doses.
M10387-14068-NLO2
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1 ATTACHMENT 3-l j Consumers Power Company
} Big Rock Point Plant Docket 50-155 ,
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i MIO487-0154A-NLO2-l 1
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Environmental Equipment Qualification regulations in 10 CFR 50.49 require that Licensees establish a program to ensure the following:
- 1. Safety-related electrical equipment located in a harsh environment are capable of performing their design basis event mitigating functions.
- 2. Nonsafety-related electrical equipment located in a harsh environment not included in the scope of qualification will not prevent satisfactory accomplishment of safety functions of safety-related equipment or mislead the operator assuming a postulated failure.
In the initial development of the EEQ list for Big Rock Point, the Control Rod Drive System Solenoids (NC-22's and NC-27's) were included on the list. The following detailed evaluation of these components and their required safety functions provides the basis for removing this equipment and the associated cabling from the EEQ list.
The function of these solenoid valves are for on/off control of instrument air to the pneumatic operators of the control valves in the reactor protection system. These solenoids are normally energized and are automatically de-energized by actuation of the reactor protection system. Following this initial de-energization, the valves remain de-energized for the post-accident period.
, A subsequent failure which re-energizes the solenoids does not cause control rod withdrawal. This failure would only close the associated scram valves and open the associated dump tank valves which, in itself, would not cause rod withdrawal. Rod withdrawal would still require operator action to select a drive and give a withdrawal signal to the control rod drive system to initiate drive movement.
Accident Response As discussed in EEQ File 4.10 for the ASCO solenoid valves which was reviewed during the September, 1986 inspection, these solenoids perform their safety function (de-energization) immediately during the start of a design basis accident. As shown on the containment pressure vs time profile (Figure 2), a step increase of 1.0 psig occurs within seconds of the start of either the large or intermediate size break accidents inside containment causing a reactor scram. The temperature, as shown on the containment temperature vs time profile (Figure 1) increases only slightly before the scram actuation signal occurs. The accident radiation would not be present during this period before the solenoid de-energization occurs. For smaller, breaks reactor i protection system actuation occurs within 100 seconds, however, as shown on 2 )
. Figures 1 and 2 for the .008 ft break, containment environment (pressure, temperature) does not become harsh until after scram has occurred, and subsequent reactor depressurization actuation. The solenoid valves also become submerged within 4 to 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> of the start of an accident, but not before the valves have performed their safety function. Therefore, the environment is not harsh during the period prior to de-energization. Accident parameters only become factors when addressing the potential failure modes of the solenoid valves after they have performed their safety function to insure that subsequent failure would not change their safety function or mislead an operator.
MIO487-0154A-NLO2 l
l
. 2 1
Failure Modes Evaluation The solenoid valves are supplied by 120V AC from the reactor protection system in the reset mode. The energized solenoids allow instrument air to be supplied to the pneumatic operators to close the scram control valves and open the scram dump tank valves. Each solenoid valve circuit originates from the trip unit of the reactor protection system, through a circuit breaker, both located in the main control room. The cable from the circuit breaker is routed outside containment to the electrical penetration, from the penetration splice inside containment to the solenoid coil and then returns the same route to the trip unit. With the solenoids energized (reset mode), operators are able to select and move individual control rods to manipulate reactor power using hydraulic water supplied from the control rod drive system.
Upon actuation of the reactor protection system, the trip unit removes the 120V AC source from the circuit causing the solenoids to drop-out to their de-energized position. This vents the instrument air from the pneumatic control valves causing them to reverse position. This action vents the top side of the CRD mechanism and supplies hydraulic pressure to the underside which initiates control rod insertion. Following insertion, the locking mechanism of the collet assembly prevents further rod withdrawal until the reactor protection system is reset (scram solenoids energized) and the operator manually initiates a withdrawal signal. Energization of the scram solenoids alone will not withdraw control rods.
Three failure modes exist with the operation of the solenoid valve circuits (valves and cable) after initial actuation. The component failure can be assumed to cause an open circuit, short circuit, or the remote occurrence of a
" hot" short wherein the de-energized solenoid valve circuit is energized from
, an independent voltage source.
If an assumed failure in either the solenoid valve coil or cable result in an open circuit, the solenoid would de-energize just as though a reactor protection system actuation had occurred. This failure would keep the solenoid valve circuit de-energized, a mode in which it performs its intended safety function. l q If an assumed failure in the cable cause a short the terminal voltage at the solenoid coil would be reduced and current flow in the circuit would increase instantly causing the circuit breaker in the reactor protection 1 system unit to trip. The resulting open circuit would de-energize the solenoid !
as above and it performs its intended safety function.
Note that with the two assumed failures above, the solenoids would de-energize to perform the safety function whether or not initial actuation had occurred.
The third assumed failure is a " hot" short in the cable causing the solenoid to re-energize. In this case a " hot" short must occur in circuits for both channels of a specific pair of scram solenoid valves in addition to a hot short causing one out of two master scram solenoids to re-energize. This would require three hot shorts to occur simultaneously for a specific set of MIO487-0154A-NLO2
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solenoids. This combination of failures woula cause one pair of scram valves to close (ie, reset). However, as discussed earlier, control rod withdrawal would not occur without additional operator action.
I Even if the long term accident environment would cause degradation of the elastomers internal to the solenoid valves, subsequent control rod withdrawal could not occur without operator action assuming that control rod drive hydraulic pressure was available.
Conclusions
]
l Based upon the above discussions, the control rod drive system solenoid valves
- and cables perform their safety function in a non-harsh environment. No significant increase in environment parameters (ie, temperature, pressure, humidity or radiation exposure) occurs until after the solenoids and their associated cables have performed their safety functions. Although the ,
i environment does become harsh after initial actuation, the failure modes analysis has shown that subsequent failure of the control rod drive solenoid schemes has no effect on accident mitigation. This concludes that the control rod drive system solenoid valves and associated cabling can justifiably be removed from the Big Rock Point EEQ program. With respect to aging of these components in their normal ambient environment, the current surveillance and preventative maintenance programs, associated with this equipment will detect and correct any operational degradation. This equipment is accessible for maintenance at any time during plant shutdowns. Current Technical Specification surveillance requirements ensure proper performance of these components and a periodic replacement interval for elastomer parts of the solenoids has been established in accordance with industry and vendor recommendations.
I.
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ATTACHMENT 4 Consumers Power Company Big Rock Point Plant Docket 50-155 JUSTIFICATIONS FOR CONTINUED OPERATION FOR COMPONENTS UTILIZING BUTYL-RUBBER AND POLYETHYLENE INSULATED CABLES April 10, 1987 k
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In accordance with the guidance of NRC generic letter 86-15, Consumers Power Company believes that the partial test data and analysis contained in the existing EEQ files for the butyl-rubber and polyethylene cables provide reasonable assurance that the cables in use will perform their intendta safety functions when called upon for accident mitigation. Consumers Power Company supports that these qualification files, in themselves, provide justification for continued plant operation until staff concerns are resolved, however, functional JCO's are also provided to show that subsequent failure of the cables under accident conditions will not result in significant degradation of any safety function or provide misleading information to the operators.
As part of developing these functional JCO's included in our March 6, 1987 submittal, a review of the Big Rock Point Emergency Procedure for " Loss of Caolant" was performed to identify any operator interface problems. This procedure already included a precautionary statement which instructs operators to take manual action to initiate an automatic function whenever an automatic action did not occur that should have occurred. This is consistent with the action requirements of the JCO's. As a result of this review, Consumers Power Company believes that subsequent failures as discussed in the JCO's do not inhibit accident mitigation nor mislead the operator in actions taken in accordance with the Emergency Procedures.
These JCO's and the potential failure of cables discussed have been reviewed with all licensed personnel and a plant " Operations Admin Memo" discussing this issue will remain in ef fect until long term resolution is completed.
- The following safety systems needed to mitigate a loss of coolant accident contain cables with the butyl-rubber and polyethylene insulation under discussion.
(
- Containment Isolation System Primary Core Spray System
- Post Incident Cooling System
- Back-up Enclosure Spray JCOs for these systems were written based upon the JCOs provided to the staff during earlier phases of the EEQ effort.
Consumers Power concludes that this information provides justification for continued operation of Big Rock Point while resolving the similarity concerns and that no unreviewed safety question exists.
MIO387-1406C-NLO2
. 2 CONTAINMENT ISOLATION Containment Isolation occurs on three monitored parameters at Big Rock Point; low reactor vessel level, high containment pressure, or loss of power. As discussed in our March 15, 1982 submittal, for large and intermediate size breaks, reactor low level is reached prior to one minute. For small breaks reactor low level is not reached until up to 45 minutes, however, the isola-tion function is actuated by the containment high pressure within 100 seconds.
Using the above, the following justifications are made with respect to the Containment Isolation schemes containing cable of polyethylene or Butyl rubber insulation construction.
- 1. The Containment Isolation paths for the containment clean and dirty sump discharge and fuel pool drain lines each contain two air operated control valves, one on each side of containment. The control valves are actuated by solenoid valves which manipulate control air to open or close the control valves. Position switches are also included on each valve.
Containment Isolation Path: Clean Sump Discharge 1 Inside Outside CV-4031 Control valve (in process path) CV-4102
- SV-4869 Solenoid valve SV-4895 POS-6624 Position switch POS-6628 Containment Isolation Path: Dirty Sump Discharge Inside Outside CV-4025 Control valve (in process path) CV-4103 SV-4891 Solenoid valve SV-4896 POS-6623 Position switch POS-6629 t
Containment Isolation Path: Feel Pool Drain Line Inside Outside CV-4027 Control valve (in process path) CV-4117 SV-4876 Solenoid valve SV-4922 POS-6618 Position switch POS-6660 These schemes are or " fail-safe" design in that loss of control air or electrical power cause the isolation valves to close.
MIO387-1406C-NLO2 l
3 The environment inside containment is assumed to cause failure of the cables for equipment inside containment. However, outside containment in the steam tunnel area where redundant equipment is located, the immediate post accident environment remains normal. Justification for continued operation for these circuits utilizing cable of butyl-rubber and/or polyethylene insulation is based on the following.
A. The equipment will have performed its safety function prior to failure. With Containment Isolation occurring in a short time period after the LOCA, the equipment sees only limited exposure to the harsh environment and failure prior to actuation is very remote.
1 B. Redundant equipment is available to substitute for a failure of the components located inside containment. If indication of inside valve closure is unavailable due to failure, operators can verify isolation from indication of outside containment valve closure.
- 2. The Containment Isolation paths for the treated waste return, demineral-ized water, and clean-up resin sluice lines each contain one air operated control valve and either a check valve or normally closed manual valve.
Containment Isolation Path: CU Demin Resin Sluice Inside Outside CV-4091,
-4092, -4093 Control valves (in process path) Manual SV-4879 Solenoid valve POS-6653,
-6654, -6655 Position switches Containment Isolation Path: Treated Waste Return Inside Outside CV-4049 Control valve (in process path) Check Valve SV-4892 Solenoid valve POS-6626 Position switch Containment Isolation Path: Demin Water Inside Outside r .
l Check Valve Control valve (in process path) CV-4105 Solenoid valve SV-4897 {
1 Position switch POS-6649 M10387-1406C-NLO2 l l
4 These schemes are of " fail safe" design in that loss of control air or
! electrical power causes the control valves to close. Additionally, the check valves and manual valve perform their safety function without electrical components.
The resin sluice line control valves and the manual isolation valve are normally closed per procedure during power operation except in the case of transferring resins. In the remote event that a break occurred during the time resins were being transferred, the qualified solenoid valves will perform their isolation function. The only portion of this scheme containing cables of polyethylene or butyl-rubber is the position indica-tion.
The treated waste return and demineralized water lines contain self-actuating check valves for performing the isolation function. Even if it is assumed that extended exposure to the harsh environment caused failure of the polyethylene or butyl-rubber cables, containment isolation is assured.
Justification for continued operation for these circuits utilizing cable of butyl-rubber and/or polyethylene insulation is based upon the following:
A. The equipment will have performed its safety function prior to failure. With containment isolation occurring in a short time period after a LOCA, the equipment sees only limited exposure to the harsh environment and failure prior to actuation is very remote. Per procedure, operators check position indication immediately following isolation actuation.
B. Redundant equipment (check valves or manual valve) not dependent on electrical operability are available to perform the isolation function. Operators are aware that these lines have containment integrity intact at all times.
- 3. The Containment Isolation path for the containment ventilation supply and exhaust air lines each contains two air operated control valves, both outside of containment. The control valves are activated by solenoid valves which manipulates control air to open or close the valve.
Position switches are also included on each valve.
Containment Isolation Path: RB Supply Air Outside 1 Outside 2 CV-4096 Control valve (in process path) CV-4092 SV-9151 Solenoid valve SV-9152 POS-9101 Position switch POS-9102 MIO387-1406C-NLO2 d
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1 f Containment Isolation Path: RB Exhaust Air j
i Outside 1 Outside 2 l CV-4094 Control valve (in process path) CV-4095 j SV-9153 Solenoid valve SV-9154
{ POS-9103 Position switch POS-9104 i
I In addition to the containment isolation function, these valves may also
! be required to reopen to accomplish the vacuum relief function during the I post-LOCA period when containment sprays are actuated to condense the
- steam pressure environment inside containment. Containment spray is j actuated at a containment pressure of 2.2 psig which as discussed in our l March 15, 1982 submittal occurs at approximately 75 seconds into the
event.
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l This equipment and the subject cables are located outside containment and
! subject to normal ambient temperature, pressure, and relative humidity i prior to and following a LOCA. Radiation exposure during the 30 day period following a LOCA will only be 2.48 x 104 rads and occurs after the components have performed their safety function.
Justification for continued operation for these circuits using cable of butyl-rubber and/or polyethylene insulation is based upon the following:
A. The equipment will have performed its safety function prior to failure. With containment isolation and vacuum-relief occurring in a i short time period after a LOCA, the equipment sees only limited l radiation exposure and a normal thermal environment and failure prior l co accomplishing these function is very remote. Valve position i indication circuits are also located outside of containment.
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, PRIMARY CORE SPRAY i
l l This system automatically actuates the primary core spray valve operators upon i receiving low reactor water level and low reactor pressure signals. Cables which are made of the subject butyl-rubber and polyethylene insulation material serve the level switches LS-RE09A through D and pressure switches 3
PS-IG11A through D. These devices are located inside the Reactor Building.
As discussed in the March 15, 1982 submittal, core spray activation occurs i within 2-5 minutes following large and intermediate size breaks. For small I breaks, core spray occurs up to 45 minutes following the event, however, due i to the limited size of the break, containment atmospheric conditions are
- significantly less than design parameters until RDS actuation. The justifica-2 tion for continued operation with the subject cables installed is based on the
! following:
I i MIO387-1406C-NLO2 i
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- 6 A. A redundant fully qualified, back-up core spray system (M0-7070 &
M0-7071) is available to substitute for the assumed failures in the primary spray system.
B. Based upon the actuation times and environment at time or actuation as discussed above, the equipment will perform its safety function prior to failure.
C. Since the suspect cables are only used in the auttaatic actuation circuits for the core spray systems the plant can be safely shut down in the absence of the cables by manually initiating primary core spray. The ability to reclose the valves is not affected by a failure in the suspect cables, an action which may be necessary when entering the recirculation mode. Information needed to access the timing /need for operator action (ie containment water level) is via circuits which do not contain butly-rubber or polyethylene cables.
REACTOR PROTECTION SYSTEM (RPS)
As discussed in our March 15, 1982 submittal, RPS actuation can occur from the following monitored parameters during LOCA events:
- Low Steam Drum Water Level
- High Containment Pressure
- Loss of Power Reactor Protection System actuation occurs very shortly following a LOCA event. For large and intermediate size breaks reactor low level is reached prior to one minute. For smaller breaks, scram is actuated by high contain-ment pressure within 100 seconds.
The following RPS component schemes use cable of polyethylene or butyl-rubber insulation:
- Main Steam Isolation Valve (M0-7050) Limit Switch
- Low Reactor Level Switches (LS RE09 A-D)
' High Containment Pressure Switches (PS-664-7)
The main steam isolation valve (M0-7050) limit switch initiates RPS actuation for load rejection transients and not LOCA events. A failure of the cable (open or short circuit) between the limit switch and RPS will not prevent a l reactor scram from occurring during a LOCA event. Justification for continued operation is based upon the fact that the plant can be safely shutdown in the absence of the cables (assumed failure) in this scheme.
The low reactor water level, and high containment pressure circuits initiate RPS actuation (scram and containment isolation) during LOCA events. In addition, the low steam drum water level schemes are the first to initiate MIO387-1406C-NLO2
7 reactor scram during LOCA events. The sensor's contact opens on an initiating signal to actuate the RPS.
During LOCA events, actuation of these circuits occurs very early. Although not discussed earlier, the low steam drum level setpoint is reached much earlier than the low reactor level setpoint. Due to these early actuation times, the cables / components in these schemes do not experience the harsh environment prior to actuation. In addition, the containment high pressure switches and cabling are located outside of containment in the Electrical Penetration Room. Justification for continued operation is based on the following:
A. The cables / components will have performed their safety functions prior to failure. RPS actuation occurs within 1 minute following a break event and with limited exposure to the environment in this short period, failure prior to actuation is very remote.
B. The plant can be safely shut down in the absence of the cables and components. Assuming failure of the cables in the sensor circuits, manual initiation of the RPS can occur in the Control Room on circuits not subjected to harsh environments.
BACKUP ENCLOSURE SPRAY The back-up enclosure spray valve is manually actuated following a failure of the primary enclosure spray valve to open. As discussed in our March 15, 1982 submittal, the need for enclosure spray occurs for large steam line breaks.
The primary enclosure spray valve actuates within 75 seconds of the event.
Should the unlikely failure of the primary enclosure spray valve occur at this time, the back-up spray would be manually actuated immediately.
Within this system, cables going to the back-up enclosure spray operator are made of the subject butyl-rubber insulation material. Our submittals dated September 19, 1986 and October 10, 1987 provide a detailed description of these spray systems and a justification for continued operation with MO-7068 EEQ documentation inadequate. Although this valve operator has been replaced with a fully qualified operator, the JC0 can be used for this cable issue as well. Thus justification for continued operation can be summarized as follows:
A. Another system (primary enclosure spray valve) is capable of providing the required safety function. Although the primary spray valve has suspect cable in its actuation circuit, it can be manually initiated to accomplish the safety function.
B. The equipment will perform its safety function prior to failure.
Since the valve is actuated shortly ascer a LOCA with limited cable exposure to the environment in this short period, failure prior to MIO387-1406C-NLO2
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- 8 actuation is unlikely. If the valve fails open after actuation the ability to furnish adequate core spray flow will not be jeopardized as discussed in the September 19, 1986 submittal.
C. The plant can be safely shut down in the absence of the cables. The unique function of the back-up enclosure spray is to provide post-accident iodine washdown. Failure to accomplish this function due to l the changes made in operating procedure EMP 3.3 as described in the September 19, 1986 letter, will not prevent safe shutdown of the plant. The EMP 3.3 changes will remain in place until the cable issue is resolved.
POST INCIDENT COOLING SYSTEM This system uses fire water to cool the containment sump water which is then recirculated by the core spray pumps back to the reactor vessel to establish long term cooling. Cables which are made of butyl-rubber and polyethylene insulation material serve a valve motor operator (MO-7066) and the core spray pump motors. All the components in this system are manually initiated from the Control Room and are all located in the Core Spray Equipment Room which is located outside of and shielded from containment. Post-accident conditions for this location are:
Radiation dose (30 day) 4.29 x 10 4R Temperature (maximum ambient) 169'F No other parameters apply to this area.
Based upon the above, continued operation is justified as follows:
A. The equipment / cables will have performed their safety function prior to failure. As discussed, the accident conditions are not very harsh l and failure of the cables under these conditions is not expected.
l B. For an assumed failure of the cables to M0-7066, a fully qualified redundant system, MO-7080, is capable of providing cooling water to the core spray heat exchanger.
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MIO387-1406C-NLO2 1