IR 05000220/1988034
| ML18038A438 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 01/10/1989 |
| From: | Anderson C, Roy Mathew, Paolino R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML18038A436 | List: |
| References | |
| CON-IIT07-432A-91, CON-IIT07-432B-91, CON-IIT07-464A-91, CON-IIT07-464B-91, CON-IIT7-431A-91, CON-IIT7-432A-91, CON-IIT7-432B-91, CON-IIT7-464-91A, CON-IIT7-464A-91, CON-IIT7-464B-91, RTR-NUREG-0737, RTR-NUREG-737, RTR-REGGD-01.097, RTR-REGGD-1.097 50-220-88-34, 50-410-88-32, GL-82-33, NUREG-1455, NUDOCS 8902020246 | |
| Download: ML18038A438 (44) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report Nos.
50-220/88-34 50-410/88-32 Docket Nos.
50-220/50-410 License Nos.
DPR-63/NPF-69 Pri ority Category C
Licensee:
Nia ara Mohawk Power Com an 301 Plainfield Road S racuse New York 13212 Facility Name:
Nine Mile Point Nuclear Station Units 1 and
Inspection At:
Scriba New York Inspection Conducted:
November 14-18 1988 Inspectors R. J'aolino, Senior Reactor Engineer/PSS'ate R.
K. Mathew, Reactor Engineer/PSS Other Partici ants and Contributors to the Re ort Include:
Allan C. Udy, NRC Contractor INEL date Ron Yander
,
NRC Contractor - INEL Approved by:
C C. J.
Anderson, Chief - Plant Systems Section EB/DRS
/ /o fT date Ins ection Summar:
Inspection on November 14-18, 1988 (Combined Inspection Reports Nos. 50-220/88-34 and 50-410/88-32)
Areas Ins ected:
A special announced inspection of representative subsystems of the post-accident monitoring instrumentation in accordance with Regulatory Guide 1.97, Revision 2, for Unit 1 and Revision 3 for Unit 2.
The inspection assessed the licensee conformance to requirements specified in the Order, dated June 12, 1984, to c'ommitments made per generic letter 82-33 and supplement 1 to NUREG-0737.
PDR ADOCK 0500 G
Results:
Of the areas, inspected there was one apparent violation with three examples of not complying with 10 CFR 50.49 requirements for safety related electrical equipment at NMP-1.
In addition, significant deficiencies were identified by the licensee and the NRC.
These deficiencies indicate that the licensee does not conform to Section 6 of Supplement 1 to NUREG-0737 regarding Regulatory Guide 1.97 "Applications to Emergency Response Facilities."
The deficiencies related to the following areas:
~
~Redundanc A number of deficiencies were identified with control room R.G.
1.97 instrumentation regarding electrical and physical separation.
Interfaces,-
A number of circuits were identified which either lacked lE/non-1E isolation devices or which contained inadequate isolation devices.
~
Ta in
& Markin
- R.G.
1.97 instrumentation in the control room was not appropriately marked.
~
Dis lay 8 Recordin Recording of instrumentation readout for one channel for certain instruments was not provided.
Specific findings for NMP-1 are identified in attachment, 1 (AEB).
For NMP-2 minor deficiencies were identified involving differences in documented instrument ranges and installed instrument configuration.
The majority of these deficiencies require changes to the FSAR and engineering drawings.
Specific deficiencies for NMP-2 are identified in attachment l(C).
S S
DETAILS 1.0 Persons Contacted Nia ara Mohawk Power Cor oration NMPC
- J
- U T.
- f
- K.
J W.
E.
J.
~ W.
J.
J.
- J T.
- E
- D
- p
- G S.
.* R.
L.
- B
- A
J K.
- p
- J
- K.
G.
Beckham L. Benson Buiva J.
Chwalek J.
Constance A. Dahlberg L. Di 1 1 on B.
Davey Dehart N. Duell P.
James, Jr.
Jirousek L. Kibbe R. Kinsley Kolceski W.
Leach Lesurdo D.
MacEwan H. Montgomery C. Nicolos J.
Pasternak Price L. Rademacher Randall D. Sassani J.
Sweet H. Snyder B.
Thomas Volza W.
Wi 1czek, Jr.
L. Willis Yackel QA/QC Nuclear Consultant Unit 2 Engineering Chem
& Rad.
Mgt, Unit
Nuclear Electrical Design Stations Manager, Operations, Unit
Supervisor QA Audits NMP2 Oper Engineering, Unit 2 Chem
& Rad Mgt, Unit 1 I&C, Unit
EQ Manager NMP2 Oper I&C Supervisor EQ Engineering Radiation Protection PSC NYSEG Regulatory Compliance I&C, Unit 2 Manager, Site Engineering EQ Engineer Director of Compliance, Unit 1 Operations, Superintendent, Unit 1 Project Engineering, Unit 2 Maintenance Superintendent, Electr, Unit 1 Site Liasion Engineering Consulting Services Rad Protection Manager Manager Nuclear Technology General Superintendent I&C, Unit
1.2 Stone and Webster En ineerin Cor oration A.
L.
A.
Issa I 1 ly Sclocchini Seismic Engineer EQ Engineer Design Engineer 1.3 U.S. Nuclear Re viator Commission C.
R.
Anderson Temps Chief, Plant Systems Section EB/DRS Resident Inspector
" denotes personnel attending exit meeting of November 18, 1988
2.0 Introduction
~Back round The purpose of this inspection was to verify the implementation of instru-mentation systems for assessing plant conditions during and following the course of an accident that meets the criteria specified in Regulatory Guide (RG) 1.97, Revision 2 for Unit 1 and Revision 3 for Unit 2.
These systems were inspected to determine if they were installed in accordance with generic letter number 82-33 "Requirements for Emergency Response Capability" (Supplement 1 to NUREG-0737).
This letter, issued on December 17, 1982, specifies those requirements regarding emergency response capabilities that have been approved by the NRC for implementation.
This supplement also discusses, in part, the application of RG 1.97 to the emergency response facilities, including the control room (CR), technical support center (TSC)
and the emergency response facility (EOF) at power plants.
Regulatory Guide 1.97 identifies the plant variables to be measured and the instrumentation criteria for assuring acceptable emergency response capabilities during and following the course of an accident.
e Regulatory Guide 1.97 divides Post Accident Instrumentation into 3 categories and 5 types.
The three design categories are noted as 1,
2 and 3.
Category
has the most stringent design requirements and category 3 has the least stringent.
The. five types of instrumentation identified in the Regulatory Guide are types A, B, C, D, and E.
Type A variables are plant specific and classified by the licensee.
Type B variables provide information to indicate that plant safety functions are being accomplished.
Type C
variables provide information regarding the breach of barriers for fission product release.
Type D variables indicate the operation of individual safety systems.
Type E variables are those that indicate and determine the magnitude of the release of radioactive materials.
Each variable type ca'n be any design category.
However, type A variables can only be design Category 1.
Cor res ondence The licensee's response to RG 1.97 for Unit 1 was provided in submittals dated April 2, 1984; October 18, 1985 and December 6,
1985.
For Unit 2 the licensee's RG 1.97 response was provided in submittals dated October 5, 1984, January 20, 1986 and March 1986.
The Safety Evaluation Report (SER) was issued by the NRC on March 3, 1986 for Unit 1 and May 5, 1986 for Unit 2.
The SERs specify staff positions for licensee deviations and exceptions from the guideline References The specific references used to assess the licensee's response to Regulatory Guide 1.97 are as identified below:
~ Regulatory Guide 1.97, Revision 2 and Revision 3 "Instrumentation for Light Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident."
~ Safety Evaluation Report - Emergency Response Capability, Conformance to Regulatory Guide 1.97, Rev.
2 & 3.
~ Niagara Mohawk Power Corporation Units 1&2 Final Safety Analysis
'eport (FSAR), Chapter 7.
~ Licensee procedures and reference drawings as shown in attachments
& 2.
~ Technical Supplement to Petition for Conversion from Provisional Operating Licenses to Full Term Operating License, Nine Mile Nuclear Station, July 1972.
~ Amendment 1 to Application to Convert Provisional Operating License to Full Term Operating License, November 1973.
Ins ection Sco e
The NRC inspection scope included:
equipment qualification (Seismic and Environmental),
redundancy of power supplies, measured variables, display and recording methods used, independence and separation of electrical circuits, range and overlapping features of multiple instrument indicators, equipment identification for RG 1.97 instruments, service, test and surveillance frequency, direct and indirect measurements of parameters of interest.
The safety related (g) and Eg master equipment lists were reviewed for the instruments selected, to ascertain whether they had been evaluated and tested to the appropriate environmental, quality assurance (gA) and seismic qualification requirements.
The gA procurement of these instruments was also reviewed.
3.0 Ins ection Details The inspectors held di scussions with various members of the licensee's staff, reviewed drawings (Attachment 2) and procedures, and selected variables for systems walkdown.
Walkdowns were performed for control room instruments to assess the implementation of the RG 1.97 Rev.
2 for Unit 1 and Rev.
3 for Unit Instrument vari abl es revi ewed for NAP-1 included reactor cool ant level, reactor pressure, drywel 1 pressure, drywel1 atmospheric temperature, containment hydrogen/oxygen concentration, s'uppression pool water level and suppression pool water temperature.
The NNP-2 inspection included primary containment pressure, coolant water flow to ESF system component, coolant water temperature to ESF system component, RHR heat exchanger outlet temperature, LPCI system flow, drywell spray flow, suppression pool pressure and primary containment area radiation.
Characteristics examined for each variable include identity, location, function, separation (physical/electrical),
isolation, seismic, power source, environmental qualification status and instrument range.
Items not conforming with Section 6 of Supplement 1 to NUREG-0737 are discussed in the following section.
4.0 Re viator.
Guide 1.97 Variables Evaluated Unit
4. 1 Reactor Pressure Regulatory Guide 1.97 classifies this as a Category 1 variable and the licensee's commitment to Section 6, Supplement 1 of NUREG-0737 resubmittal dated October 5, 1987 has specified that this variable is a Type A variable.
The monitoring of this variable is accompli shed by using two pressure sensing channels with indication in the control room.
One channel has a recorder.
The variable is also monitored by the Safety Parameter Display System (SPDS).
The instrumentation for this variable does not meet the Category 1 criteria specified in Section 6, Supplement 1 to NUREG-0737 regarding Regulatory Guide 1.97 requirements for separation.
Section 1.3. 1.b of the Regulatory Guide states, in part, that redundant or diverse channels should be electrically independent and physically separated from each other.
The pressure recorder PR-ID75 interfaces with PT36-31 and PT36-32 through switch no.
ID20A.
No isolation device or separation is provided between channels.
Also, all power sources, numbers ACV-P, ACV-R, and ACV-S normally are powered by RPS Bus ll, Circuit 12.
Therefore, they are not considered independent, redundant power sources.
This apparent deviation from Section 6 of Supplement 1 to NUREG-0737 requirements regarding Regulatory Guide 1.97 is an unresolved item pending the licensee developing the technical basis for the variations from R.G.
1.97 and NRC review of this information.
(88-220/88-34-01)
4.2 Dr well Atmos heric Tem erature Regulatory Guide 1.97 generally classifies the drywell atmospheric temperature as a Category 2 variable.
However, the licensee has specified that this variable is a Type A variable.
Because it is a
Type A variable, the instrumentation supplied should be Category l.
The monitoring of this variable is accomplished by three channels of instrumentation.
The instrumentation for this variable does not meet the Category 1 guidance specified in RG 1.97.
Section 1.3. 1(b) of
th
4.3 the Regulatory Guide states, in part, that:
"redundant or diverse channels should be electrically independent and physically separated from each other...
At least one channel should be displayed on a
direct indicating or recording device."
All of these instrument channels (Nos.
TE201-36, TE201-27 and TE201-33)
have a
common power supply.
Therefore, the channels were not provided with complete independence and redundancy.
No recorder for readout data is provided.
This apparant deviation from Section 6 of Supplement 1 to NUREG-0737 regarding compliance with Regulatory Guide 1.97 criteria is an unresolved item pending the licensee developing the technical basis for the variations from R.G, 1.97 and NRC review of this information.
(50-220/88-34-02)
In addition, environmental qualification for the three instrument channel temperature sensing elements (TE201-36A, TE-201-50A and TE201-51A)
has not been established.
This is a violation of 10 CFR 50.49 (b)(3) and (F) which requires that these items of electrical equipment be qualified.
(50-220/88-34-03)
Su ression Pool Water Level Regulatory Guide 1.97 classifies suppression pool water level as a
Category 1 variable.
The licensee's resubmittal commitment to Section
of Supplement 1 to NUREG-0737 dated October 5, 1987 has specified that this.variable is a Type A variable.
The monitoring of this variable is accomplished by three instrument channels ( LT58-04, LT58-05 5 LT58-06).
The instrumentation for this variable does not fully meet the Category 1 criteria specified in RG 1.97.
The level transmitter LT58-04 was not EQ listed and environmental qualification of this item has not been established.
This is in violation of
CFR 50.49(b)(3)
and (F) which requires that each item of electrical equipment important to safety be qualified.
(50-220/88-34-04)
Su ression Pool Water Tem erature Regulatory Guide 1.97 classifies suppression pool water temperature as a Category 2 variable.
However, the licensee has specified that this variable is a Type A variable.
Therefore, the instrumentation should conform to Category 1 criteria.
The monitoring of this variable is accomplished by two channels of instrumentation (TT201.2-517 and TT201.2-518).
The instrumentation for this vari'able does not fully meet the Cat'egory 1 criteria specified in RG 1.97.
The thermocouples (TE201.2-491, TE201.2-492)
and transmitters (TT201.2-517, TT201.2-518)
did not have supporting documentation to verify that they are environ-mentally qualified.
This item is in violation of 10 CFR 50.49(b)(3)
and (F) which requires that each item of electrical equipment important to safety be qualified.
(50-220/88-34-05)
0
Isolation Devices Where a Category 1 signal is used as input to a non-Category 1 system, Regulatory Guide 1.97 specifies the use of isolation devices that are fully qualified for use in Category 1 circuits.
The inspectors examined interconnecting diagrams for torus level (3), drywell pressure (4)
and reactor coolant (5) instrumentation circuits'ix out of twelve circuits examined did not use any isolation devices.
The six instru-ment channels include:
level transmitter nos.
LT58-04, LT58-05 and LT58-06, pressure transmitter nos.
PT201.2-483, PT201.2-484 and reactor coolant level transmitter LT36-33.
Typically, the inspectors found that either there were no isolation devices or that signals are directed to the plant computer by way of a dropping resistor, and in some cases, medium impedance line input resistors.
This does not conform to the Regulatory Guide 1.97, Category 1 criteria for isolation.
This apparent deviation from the Regulatory Guide 1.97 criteria is an unresolved item pending the licensee developing the technical basis for variations from R.G.
1.97 and NRC review of this information.
(50-220/88-34-06)
One isolation device (RIS-SC326)
was identified for which documenta-tion was not available to establish performance characteristics or seismic requirements.
This item is unresolved pending NRC review of licensee evaluation and corrective action.
(50-220/88-34-07)
5.0 Re viator Guide 1.97 Variables Evaluated Unit 2 5.1 Reactor Coolant Level Regulatory Guide 1.97 classifies reactor coolant level as a Category
variable.
The licensee has specified that this variable is a Type A variable.
The monitoring of this variable is accomplished by two channels of instrumentation (2ISC*LT13A 5 2ISC"LT13B) for wide range level and two channels of instrumentation (2ISC*LT9A 5 2ISC*LT9B) for the fuel zone level.
In addition to providing indicators in the reactor control room, the variable is recorded and displayed on demand on the Safety Parameter Display System (SPDS).
During the inspection of the instrumentation (2ISC*LI13A 5 2ISC*LI13B)
in the control room for this variable, the range of the variable was found to be -165 to +35 inches of water, whereas the range specified by the FSAR is 230.69 to 430.69 inches of water.
The licensee reviewed the design and determined that the range of 200" of water (-165" to
+35") is required.
,The documentation will be changed to reflect actual field conditions.
Other than the documentation correction, the instrumentation for this variable meets the criteria of RG 1.97 for Category 1 instrumentation and is acceptable.
This item is unresolved pending licensee document changes to reflect actual field conditions.
(50-410/88-32-01)
5.2 Coolin Water Flow to En ineered Safet Features ESF S stem
~Com onents Regulatory Guide 1.97 classifies cooling water flow to ESF system components as a Category 2 variable.
Since it is a Category 2 variable, the instrumentation does not require seismic qualification.
The monitoring of this variable is accomplished by two channels of instru-mentation.
The inspection for this variable noted that the range of the 2SWP'FI13A and 2SWP"FI13B channels is 0 to 8,000 gpm whereas the FSAR specifies 0 to 10,000 gpm.
For the 2SWP*FI76A and B channels it was noted that the range is 0-1,400 gpm, whereas the FSAR specifies 0-860 gpm.
It was also determined that the range for the 2SWP"FI535 channel should have been 0 to 1,000 gpm.
The licensee plans to submit an FSAR change to reflect the correct instrument range.
In addition to revising the FSAR to correct the difference in range, the licensee must determine whether the range of 0 to 8,000 gpm for the 2SWP*FI13A and 2SWP*FI13B channel meets or exceeds the 0 to 110 percent design flow range recommended by RG 1.97.
The above items are unresolved.
(50-410/88-32-02)
s With the exception of the range deficiencies, this instrumentation meets the Category 2 recommendations specified in RG 1.97 and is acceptable.
5.3 Low Pressure Coolant Injection LPCI S stem Flow Regulatory Guide. 1.97 classifies LPCI system flow as a Category
variable.
Since it is a Category 2 variable, the instrumentation does not require seismic qualification.
The monitoring of this variable is accomplished by three channels of instrumentation.
The inspection of the instrumentation (2SWP*FI14A, 2RHS*FI14B 8 C) for
'this variable noted that the indication range is 0 to 10,000 gpm; whereas, the FSAR and the FSAR question for this variable indicates ranges of 0 to 8,400 gpm and 0 to 8,000 gpm respectively.
The licensee has committed to correct the differences in the ranges specified in the FSAR and will determine whether the 0 to 10,000 gpm range is the recommended zero to 110 percent design flow range specified by RG 1.97.
With the exception of the determination of zero to 110 percent design flow for this variable, the instrumentation meets the recommendations of RG 1.97 and is acceptable.
This item is unresolved pending licensee amendment of the FSAR and evaluation and determination on meeting 0-110 percent design flow range.
(50-410/88-32-03)
l
'
0
5.4 Dr well S ra Flow
Regulatory Guide 1.97 classifies drywell spray flow as a Category
variable.
Being a Category 2 variable, the instrumentation does not requi re seismic qualification.
The monitoring of this variable is accomplished by two channels of instrumentation (2RHS*FT63A 8 2RHS"FT63B).
The inspection of the instrumentation (2RHS"FI63A 8 2RHS~FI63B) for this variable noted that the indication range is 0 to 8,000 gpm; whereas, the FSAR specifies a range of 0 to 7950 gpm.
The licensee has committed to correct the difference noted.
The instrumentation meets the recommendations of RG 1.97 and is acceptable.
6.0 Ph sical Malkdown The inspectors examined the Unit 1 and 2 control rooms to determine agree-ment with Regulatory Guide 1.97 guidance for electrical/physical separation, identification and adequacy of the instrumentation calibration program.
For Unit 2, post-accident monitoring instruments designated as Type A, B
and C and Categories 1 and 2 were specifically identified on the control panel (red outline)
so that the operator would easily discern the intended use under accident conditions.
Location and accessibility provides easy access for administrative control of all set point adjustments, module calibration and test points.
The inspectors determined that the Unit 2 control room was in conformance with the Regulatory Guide 1.97, Revision
guidance for those portions examined.
For Unit 1, the inspectors guestioned the adequacy of electrical and physical separation of the control room post-accident monitoring components.
Redundant channels were physically mounted side by side with electrical wiring from both divisions in a common bundle.
This apparent deviation from Section 6 of Supplement 1 to NUREG-0736 requirement regarding Regulatory Guide 1.97 is an unresolved item pending the licensee developing a technical basis for variation from R.G.
1.97 and NRC review
~
of this information.
(50-220/88-34-08)
In addition, the Types A, B and C post-accident instrumentation on the control boards was not specifically identified.
Regulatory Guide 1.97 Revision 2, section 1.4(B) states, in part, that:
"the instruments'esig-nated as types A,
B and C and categories 1 and 2 should be specifically identified on the control panels...."
During this inspection the licensee issued a problem report and operator aid form (88-56) for temporary identi-fication using color dots on accident assessment instrumentation.
This apparent deviation from Regulatory Guide 1.97 is an unresolved item pending the licensee developing a technical basis for variations from R.G.
1.97 and NRC review of this information.
(50-220/88-34-09)
7.0 Survei 1 lance Testin and Cal ibration Nine Mile Point, Units No.
1 and 2, employ a computerized data base for instrument calibration.
Part of the function of this data base is to generate the repetitive maintenance task orders for recalibration of instruments and to provide a historical record of instrument calibrations.
The inspectors reviewed the data base for the instruments inspected, the frequency of calibration, and the date that calibration is next due.
Verification was made to determine that procedures are in place for the performance of the calibration.
The maintenance and calibration data for Nine Mile Point, Units No.
1 and 2, instrumentation for the inspected variables are found in Attachment 3.
The inspectors noted that Unit 1 is in a long term shutdown condition, and the calibration frequency is not a
factor in regards to technical specifications.
No abnormalities were noted.
8.0 Unresolved Items Unresolved items are matters for which more information is required in order to ascertain whether they are acceptable, violations, or deviations.
Unresolved items are discussed in Sections 4. 1, 4.2, 4.5, 5. 1, 5.2, 5.3, and 6.0 of this report.
The inspectors met with licensee representatives (denoted in Details, paragraph 1) on November 18, 1988 and discussed the findings for Units No.
1 and 2.
For Unit 1 significant items discussed include channel separation, use of isolation devices, tagging and identification of post-accident monitoring instrumentation in the control room, and equipment qualification for the instrumentation inspected.
For Unit 2, significant items include:
actual instrument ranges versus documented ranges.
10.0 Follow-U Technical Meetin A meeting was held with the licensee to discuss the results of this inspection of R.G.
1.97 activities as it pertains to Unit 1.
The meeting was held on December 20, 1988 at NRC/NRR offices in Rockville, Maryland.
The licensee, NRR and Region I personnel participated in this meeting.
The primary purpose of this meeting was to discuss the reason for the numerous deviations from R.G.
1.97 noted during this inspection.
The licensee explained that it was always their intention to limit their R.G.
1.97 implementation for Unit 1 to specific criteria delineated in Section 6.2 in Supplement 1 to NUREG-0737.
They stated that the scope of their R.G.
1.97 implementation was identified in their letter dated
l
October 18, 1985 from C.
V. Mangan of Niagara Mohawk to Domenick B. Vassalo of the NRC.
They noted that they had not addressed other criteria identified in R.G.
1.97 (Rev. 2), since they were not explicity called out in Supple-ment 1 to NUREG-0737.
The licensee is currently performing a re-review of Unit 1 to the R.G.
1.97 criteria.
Another meeting is scheduled with the licensee on February 3,
1989 to discuss the results of this re-review and licensee plans to address proposed plant modifications associated with some of the R.G.
1.97 deficiencies at Unit ATTACHMENT 1 Results A.
Uiolation
CFR 50.49(b)(3)
and ( F) requires that each item of safety-related electrical equipment be qualified.
Contrary to the above, the environmental qualification for the following items was not established at the time of this inspection.
Descri iion
~Para ra h
Docket No.
Identification a. Drywell Atmosphere Sensing Elements 4.2 Nos.
TE201-36A, TE201-SOA and TE201-51A 50-220/88-34-03 b. Suppression Pool Water Level Transmitter LT58-04 4.3 50-220/88"34-04 c. Suppression Pool Water Temperature 4.4 Sensing Elements Nos.
TE201.2-491, TE201. 2-492, TE201. 2-517 and TE201. 2-518 50-220/88-34-05 B.
Unit 1 Unresolved Items Descri tion Para<apra h
Docket No.
Identification 1.
Reactor pressure channel recorder PR-ID75 interfaces with switch IO-20A common to redundant channels PT36-31
PT36-32.
No isolation or separation provided 2.
Torus level, drywel l pressure and reactor coolant electrical circuits without isolation devices 4.1 4.5 50-220/88-34-01 50-220/88-34-06 3.
Redundant accident assessment instrumentation in control room does not comply with electrical/physical separation guidance 6.0 50-220/88-34-08
I
Descri tion 4.
Drywell atmosphere temperature sensing elements TE201-36A, TE201-50A and TE201-51A have common power supply and no recorder 5.
Type A, B & C Post Accident variables not specifically identified on control boards 6. Circuit Isolation Device RI-SC326
~Para ra h
6.0 Docket No.
Identifi'cation 50-220/88-34-02 50-220/88-34-09 50-220/88-34-07 C.
Unit 2 Unresolved Items 1)
Vessel level fuel zone (Device No. 2ISC*LT13A,B) FSAR question 421.36-1 indicates range of 230.69 to 430.69.
Actual field installed indication is -165 to +35 inches WC.
No listing in Table 7.5-1 of FSAR.
Paragraph 5.1 (50-410/88-32-01)
2)
Service water to diesel generator - Div. I & II (Device No.
2SWP"FI76A&B) FSAR question 421.36-1 indicates a 0-860 gpm range.
Drawing C071M indicates a 0-1400 gpm.
No listing in FSAR Table
7.5-1.
Actual installed field indication 0-1400 x 10 gpm.
ESF cooling water flow (Device No.
2SWP FI13A&B) FSAR question 421.36-1 indicates a 0-10,000 gpm range.
FSAR Table 7.5-1 lists a 0-8000 gpm range.
Actual installed field indication is 0-8000 gpm.
Service water to diesel generator Div. III (Device No. 2SWP*FI535)
FSAR question 421.36-1 indicates a 0-650 gpm range.
Owg.
C071M indicates a 0-1000 gpm range.
No listing in FSAR Table 7.5-1.
Actual installed field indication 0-1000 x 10 gpm.
Paragraph 5.2 (50-410/88-32-02)
3)
2RHS"FI14A,B&C) FSAR question 421.36-1 indicates range of 0-8400 gpm.
General Electric dwg.
and FSAR Table 7.5-1 indicate a range of 0-10,000 gpm.
Actual installed field indicator is 0-8000 gpm.
Paragraph 5.3 (50-410/88-32-03)
D.
E.
F.
~
ATTACHMENT 2 INSTRUMENT SCHEMATICS AND DRAWINGS I.
UNIT NO.
Reactor Coolant Level C-22005-C, E21.5 Sheets 1 through
C-18015-C, S18.8 C-22004-C, E21.5 Sheets 1 and
C-34853-C, E21 Sheet
C-34830, Sheet
C-34831, Sheet
C-23087-C C-23089-C Reactor Pressure C-22004-C, E21.5, Sheet
C-23077-C, Sheets 1, 2, 3, 6, and
C-18015-C, S18.8 Dr well Pressure C-22020-C, E21.5 Sheets 1, 2, and
C-18012 Sheet
C-18014 Sheet
C-22385, 16A and 17A C-.22005-C, E21.5 Sheets 1, 5, 8, and
Dr well Atmos here Tem erature C-22020-C, E21.5 Sheets 1 and
C-18014 Sheet
E.
Suppression Chamber Water Level (Torus)
Su resion Chamber Water Level Torus C-22020-C, E21.5 Sheet
C-22015-C, E21.5 Sheet
C-18007-C Sheets 1 and
Containment H dro en/Ox en Concentration C-27003-C, E21 Sheet
C-27004-C, E21 Sheet
C-22020-C, E21.5 Sheets 8 and
C-26939-C, S18.9 C-26949-C, S18.9 Su ression Pool Water Tem erature
'I
C-18014-C, E21 Sheet
C-34853-C, E21 Sheets 1 through
,C-34854-C, E21 Sheets 1 through
II'NIT NO.
Reactor Vessel Level - Wide Ran e and Fuel Zone PID 28B PID 28C NSSS 16.020-5003 16.130-001-055 LCR IL2ISC-068 7.241-001-030 7.510-001-183 7.510-001-231 7.510-001-234 7.510-001-249 7.510-001-251 807 E 171TY Sheet
EE-3CS EK-401V LOOP 2ISC*13 Reactor Pressure PID 28A LCR IL2ISC-050 7.510-001-250 7.510-001-240 NSSS 16.020-5003F 16.130-001-055 7.241-001-030 807 E '152TY Sheets 7.510-001-251 EE-3CS 7.520-001-447 7.520-001-422 7.242-001-009 EK-401V 14 and
FSK27-19B FSK27-19C 16.130-001-054-16.130-001-056 LCR IL2ISC-069 7.241-001-011 7.510-001-187 7.510-001-233 7.510-001-240 7.510-001-250 7.241-001-006 EE-11C EK-401Y LOOP 2ISC*9 FSK 27-19A LCR IL2ISC-051 7.212-001-057 EE-4R 16.130"001-054 16.130-001-056 7.241-001-006 7.510-001-249 EE-11C 7.212-001-039 7.520-001-454 7.242-001-008 EK-401Y LOOP 2ISC*6 Containment Atmosphere H2/02 Concentration PID 82A FSK 33-02B LSK 33-02C LCR IL2CMS-143 LCR IL2CMS-147 7.510-413-281 FSK 33-02A LSK 33"02A LCR IL2CMS-142 LCR TL2CMS-146 7.510-413-281C 7.510-413-273
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0
7.510-413-274 7.510-413-276 7.510-413-278 SPEC C001C LOOP 2CMS-006 Sheets EE-460CN DP-384B LOOP 2CMS*071 1 and
7.510-413-275 7.510-413-277 7.510-413-279 CALC 12177-CS-CMS*08 EE-3HJ EE-460CQ DP-384AR CALC 12177-CS-CMS*07 Dr well Pressure/
Primar Containment Pressure 12177-CS-CMS "06 FSK 33-02A LCR IL2CMS-139 LSK 33-02E EE-3AE EE-3AL 7.159-401-507 7.159-401-509 7.159-401-126 7.159"401-144 EK-401AB BK-16HV 7.131-400-004 7.131-400-005 FSK 33D EE-4J LOOP 2CMS" 1 LOOP 2CMS"2 7.159-401-086 Su ression Pool Water Tem erature PID 82B FSK 02D LCR IL2CMS-104 LOOP 2CMS*67 LOOP 2CMS*68 LOOP 2CMS~69 LOOP 2CMS*70 Su ression Chamber Pressure LOOP 2CMS*7 Su ression Pool Water Level LOOP 2CMS*9 Dr well Atmos here Tem erature PID 82A FSK 33-02B LCR IL2CMS-140 EE-11C EE-3Q EE-3RC 7 '59-401-508 7.159-401-124 7.159-401-085 EK-401AD LOOP 2CMS~1 SPEC C071M 7.131-400-003 LCR IL2CMS-008 D.
LCR IL2CMS-009 LSK 33-02D EK-401Y 7.159-401-145 FSK 33-02C FSK 33-02K 12177-LSK-33-2K 12177-FL2CMS-106 LOOP 2CMSA174 LOOP 2CMS*175 LOOP 2CMS*11
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,C Attachment
2 CMS~101 2 CMS*130 I.
Status of Standb Power 12177-TL 2BYS-002 12177-ES K-8BYS13 J.
LPCI/RHR Flow 12177-TL2RHS-009 K.
Dr we 1 1 S ra Fl ow 12177-TL2RHS-062 2 CMS*116 2 CMS~140 12177-ES K-7CEC19 12177-TL2RHS-010 L.
RHR Heat Exchan er Outlet Tem erature EE114 M.
Coolin Water Flow to ESF Com onents C071-MAX C071-MA2 12177-EE-3AE-9 12177-EE-11E-8 12177-TL2SWP-131 12177-EE-3AL-6 12177-EE-11W 807E170TY SH6 and
12177-TLRSWP-002 12177-TL2SWP-046 12177-TL2SWP-048 7 '59-401-478 7.159-401-488 N.
Coolin Water Tem erature to ESF S stem Com onents 12177-TL2SWP-401 12177. TL2SWP-002 12177-EE-3D-4 12177-EE-3AE-9 12177-EE-1C-5 12177-EE-3AL 12177-EE-11C-5 7.159-401-173 7.159-401-120 7,159"401-506 7.159-401-122 7.159-401-508 213AT/232
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ATTACHMENT 3 MAINTENANCE AND CALIBRATION DATA I.
NINE MILE POINT UNIT NO.
1.
Reactor Coolant Level Variable Fre uenc Last Ca 1 ibration Date Next Calibration Date 2..
Reactor Pressure LT 36-03A LT 36-03D LT 36-33 LT 36-24A LT 36-24B 18m 18m 18m 18m 18m 3/17/88 3/10/88 5/7/88 1/29/88 1/29/88'utage
Outage
Outage
Outage
Outage
3.
Dr well Pressure PT 36-31 18m PT 36-32 18m PT 201.2-105 18m PT 201.2-106 18m PT 201.2-483 Monthly PT 201.2-484 Monthly 7/12/88 7/12/88 10/27/86 10/27/86 7/11/88 7/11/88 Outage
Outage
Outage
Outage
Outage
Outage
4.
Dr well Atmos heric Tem erature TE 201-36A 18m TE 201-50A 18m TE 201-51A 18m 5.
Su ression Chamber Mater Level LT 58-04 6m LT 58-05 6m LT 58-06 6m 6.
Containment H dro en/Ox en Concentration 5/17/86 5/17/86 5/17/86 11/10/88 11/10/88 11/10/88 Outage
Outage
Outage
5/5/98 5/5/89 5/5/89 201.2-217 quarterly 1/15/88 201.2-218 quarterly 12/2/87 201.2-330 quarterly 1/15/88 201.2-518 quarterly 12/2/87 Outage
Outage
Outage
Outage
y'
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0
7.
Su ression Pool Mater Tem erature TE 201.2-517 18m TE 201.2-518 18m 1/29/88 1/29/88 Outage
Outage
1.
Reactor Coolant Level Variable Fre uenc II.
NINE MILE POINT, UNIT NO.
Last Next Calibration Calibration Date
'ate 2.
Reactor Pressure LT 13A 18m LT 13B 18m LT 9A 18m LT 9B 18m 4/14/87 4/14/87 5/12/88 5/12/88 3/1/89 3/1/89 3/30/90 3/30/90 3.
Dr well Pressure PT 6A PT 6B PT 2A PT 2B 18m 18m 18m 18m 10/14/88 10/14/88 4/6/88 4/6/88 9/01/90 9/01/90 2/21/89 2/21/89 4.
Dr well Atmos heric Tem erature TE 116 TE 117 TE 118 TE 119 TE 120 TE 121 TE 122 TE 123 TE 124 18m 18m 18m 18m 18m 18m 18m 18m 18m 4/7/87 4/7/87 4/7/87 4/7/87 4/7/87 4/7/87 4/7/87 4/7/87 4/7/87 2/22/89 2/22/89 2/22/89 2/22/89 2/22/89 2/22/89 2/22/89 2/22/89 2/22/89 5.
Containment H dro en/Ox en Concentration AIT 71A Quarterly 8/20/88 AIT 71B Quar terly 8/20/88 AIT 6A Quarterly 8/26/88 AIT 6B Quarterly 8/26/88 12/8/88 12/8/88 12/13/88 12/13/88
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6.
Su ression Pool Water Level LT 9A LT 9B LT 11A LT 11B 7.
Su ression Pool Water Tem erature TE 67A TE 68A TE 69A TE 70A TE 67B TE 68B
18m 18m 18m-18m 18m 18m 18m 18m 18m 18m 18m 18m 10/1/88 10/1/88 9/22/88 9/22/88 9/21/88 9/21/88 9/21/88 9/21/88 9/21/88 9/21/88 9/21/88 9/21/88 8/19/90 8/19/90 8/10/90 8/10/90 8/9/90 8/9/90 8/9/90 8/9/90 8/9/90 8/9/90 8/9/90 8/9/90 PT 1A PT 1B 9.
Coolin Water Flow to ESF S stem Com onents FT 13A 18m FT 13B 18m FT 76A 18m FT 76B 18m FT 535 18m 4/7/87 4/7/87 8/29/87 8/29/87 1/11/88 1/11/88 1/16/88 2/22/89 2/22/89 7/16/89 7/16/89 11/30/89 11/30/89 12/3/89 10.
Coolin Water Tem erature to ESF S stem Com onents TE 31A 18m TE 31B 18m ll.
RHR Heat Exchan er Outlet Tem erature TE 13A 18m TE 13B 18m 12.
LPCI S
stem Flow 9/7/87 9/12/87 11/2/87 11/2/87 7/25/89 7/30/89 9/19/89 9/19/89 13.
Dr wel 1 S ra Flow FT 14A FT 14B FT 14C FT 63A FT 63B 18m 18m 18m 18m 18m 4/27/88 4/3/88 4/3/88 8/24/87 8/24/87 3/01/90 2/19/90 2/19/90 7/11/89 7/11/89
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14.
Primar Containment Area Radiation RE lA RE 1B RE 1C RE 1D 15.
Su ression Pool Pressure PT 17A PT 17B 18m 18m 18m 18m 18m 18m 10/19/88 10/19/88 10/19/88 10/19/88 1/16/88 1/16/88 4/22/90 4/22/90 4/22/90 4/22/90 12/3/89 12/3/89