Forwards addl info & justification to support resolution ... · qualification using methods other than strict application of data obtained from type testing in accordance with either

DOCKET NO. 50-263.. .MONTICELLO ..NSPC

Addl Info and Justification to Support resolution of enviro qualification of safety related electrical equipment. NUREG-0588

Rec'd w/ltr 4/12/83...8304220426

RECORDS FACILITY BRANCH

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DEADLINE RETURN DATE

Director of NRR Monticello Nuclear Generating Plant April 12, 1983

Attachment A

Page A-1

Generic Item Number I

"It is noted that the Licensee's methodology and analysis for aging and qualified life assessment are not technically valid."

Response to Generic Item Number I

NRC IE Bulletin 79-01B established the DOR Guidelines as the criteria for evaluating the qualification and associated documentation of safety-related electrical equipment. The objective of these guidelines is to identify Class IE equipment whose documentation does not provide reasonable assurance of environmental qualification.

The DOR Guidelines are not intended to provide for the implementation of either the 1971 or 1974 version of IEEE Standard 323 for operating reactors. The intent of the Guidelines is to provide a basis for judgements required to confirm that operating reactors are in compliance with General Design Criterion 4, specified in Appendix A of 10CFR50. General Design Criterion 4 states in part that "structures, systems, and components important to safety shall be designed to accommodate the affects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents." Thus it is clear that aging qualification using methods other than strict application of data obtained from type testing in accordance with either IEEE 323-1971 or 323-1974, and in particular accelerated thermal aging testing of 323-1974, does not necessarily imply that the equipment is unqualified to the DOR Guidelines.

The DOR Guidelines state that type testing of equipment identical in design and construction to the installed component is the preferred qualification method, although alternatives are acceptable provided justification is presented. Appendix C of the DOR Guidelines contains partial listing of materials which may be found in nuclear power plants, along with an indication of the material susceptability to thermal and radiation aging.' Other data sources may be used by the licensee in evaluating the radiation and thermal aging qualification of equipment and materials.

Appendix C of the DOR Guidelines shows the generic nature of materials categories. Categories include Nylon, Silicone Rubber, Polyethylene, EPR, Crosslinked Polyethylene, etc. It is clear that by providing these generic categories, the NRC is indicating the level of detail by which qualification can be established, and that a more detailed review is at the licensee's discretion.

The calculation files referenced in our November 1981 SER response provide documentation of the detailed analysis used to establish the qualified (thermal aging) life for equipment items. The methodology and sequence used to assess thermal aging qualification and establish a qualified life are outlined below.

Page A-2

1. Identify the specific component materials susceptible to degradation from thermal aging. Metallic components were assumed not susceptible to degradation from thermal aging. Where the specific material of construction could not be identified by the manufacturer, the most conservative data for typical materials were assumed.

2. Identify thermal aging properties of the susceptible materials. Data sources included:

a. Appendix C of the DOR Guidelines

b. Material Manufacturer Information

c. EPRI Report NP-1558

d. Test Lab Data

e. Other Sources

In each case, the most applicable data available were used with respect to both the tested geometry (bulk, thin film, o-ring, etc.) and the physical property of the material pertinent to the function of the component. For example, the critical physical property for a diaphragm would be retention of elongation, flexure strength, or tensile stength, as appropriate. The aging data utilized incorporate conservative definitions of material failure relative to retention of the critical property.

3. Determine a qualified life for thermal aging. Safety-related equipment at Monticello can be classified into three categories, based on the availability test data:

Category A: Equipment tested under design-basis-event (DBE)/HELB conditions with pre-aging included in the test program.

Category B: Equipment tested under DBE/HELB conditions without pre-aging in the test program.

Category C: Equipment not tested under DBE/HELB conditions.

Category A includes the equipment for which type tests have been conducted. These type tests include (1) simulated aging tests, including thermal and radiation aging, and (2) design-basis-event tests to verify the equipment's functional capability following simulated aging. In these cases, a qualified-life estimate was based on the

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Page A-3

thermal aging tests, activation energies for the materials of construction, and the critical physical property for these materials.

For equipment in Category B, pre-aging was not performed as part of the test sequence. Since significant margins often exist between the required environmental profiles and the tested environmental profiles, such "overtesting" can be reconfigured to account for in-service aging using the Arrhenius method.

Finally, Category C includes the few pieces of equipment for which qualification has been demonstrated by analysis only. This equipment is exposed to a limited number of harsh-environment parameters, which are usually lower than they would be for Category A or B. Aging analysis for these components is based on use of the Arrhenius method to compare applicable materials test data with normal plant conditions to establish a qualified life.

In the absence of a pre-accident aging test for an equipment item, the accident "overtesting" provides the next best measurement of a qualified thermal aging life. This conclusion is based on the following considerations:

The materials used in the test specimen provide the most accurate indication of how the materials in the installed equipment will age. The materials used in the test specimen are in most cases identical to the materials used in the installed equipment. Where differences exist, the tested materials usually meet the same functional design specifications.

The atmosphere in the test chamber is acceptably similar to normal plant conditions to simulate normal aging. No accelerated aging test environment can accurately simulate all aspects of normal plant aging. Air oven aging is the most widely used accelerated aging method. However, this method results in an extremely dry environment which excludes the considerable aging effects of moisture. The test chamber atmosphere during LOCA/HELB accident testing also has disadvantages, in that it has a lower than normal concentration of oxygen. Oxygen is present in the chamber in some quantities at the start of testing and, as chamber temperature rises, gases (including oxygen) are driven out of solution from water present in the chamber in the form of spray solution or a vapor source pool. This oxygen is available to participate in aging reactions. Thus, test chamber conditions are adequate to provide reasonable assurance of the thermal aging capability of the equipment being tested.

Page A-4

The Arrhenius method predicts the effects of normal aging mechanisms by modeling them as chemical reactions. The most rapid reaction rate produces the most significant aging mechanism and therefore the lowest activation energy. Since most activation energies are experimentally derived from air oven aging tests, they predict the most rapid degradation reaction rate exhibited by the material being tested (usually oxidation). Use of this activation energy to extrapolate accident simulation testing yields a significant benefit. If oxidation is the controlling aging mechanism, then life estimates extrapolated from accident test data are biased towards this mechanism by use of the lower activation energy associated with oxidation. However, if a moisture-induced aging mechanism is predominant, the test data will reflect resulting failures. In either case, use of accident test data leads to conservative estimates of equipment aging capability because such testing does not usually take the equipment to failure.

Additional conservatism is included in all Arrhenius calculations in the following manner:

a. The lowest listed apparent activation energy for a material was used unless a justification for a higher value was provided.

b. A conservatively high normal operating temperature was used to calculate a qualified life.

c. A conservative activation energy of 0.5 eV was assumed for materials where adequate aging data was unavailable.

d. A conservative high material temperature was assumed when evaluating equipment items that generate heat, such as energized solenoids and running motors.

4. Evaluate effect of material degradation. When the qualified life for a component material was less than required, the effect of the material failure was evaluated. If it was determined that failure of this material or subcomponent would not affect the operability of the component, this material was no longer considered the limiting material, and the next most limiting material was used to establish the qualified life for the equipment item. An example of this situation would be a cover gasket on a component not subject to harsh temperature, pressure, or relative humidity conditions. In this situation, failure of the cover gasket due to age-related degradation would not have a deleterious effect on the component's functionability.

, 0 1

Page A-5

For equipment exposed to small accident temperature transients and/or durations, the normal age-related degradation will not be significantly accelerated during post-accident conditions. Therefore, catastrophic failure due to thermal aging under these circumstances is very unlikely.

The November response to the June 1981 SER previously outlined this methodology for evaluating the qualified life of safety-related electrical equipment. The methodology, described above in greater detail, complies with the NRC requirements established in IE Bulletin 79-01B to provide reasonable assurance that the equipment will perform its safety function when required.

The same methodology was evaluated by the FRC in Section 4.3.6 of the TER. The result of that evaluation "concluded that the Licensee has provided a satisfactory response" to aging concerns.

Generic Item Number 2

The specific analyses documenting qualification for various environmental parameters were not included in the documentation provided to the FRC during the TER review.

Response to Generic Item Number 2

Calculation Files were initially established to document, in detail, equipment qualification during the response effort to IEB 79-01B. These files were subsequently revised as new and better information was obtained. During the effort to respond to the June 1981 SER, separate calculation files were developed solely to respond to NRC concerns about particular deficiencies. These files were general in nature, and often referenced the detailed calculation files produced for the 79-01B Response. In their request for supplemental documentation, the FRC requested several of these SER files without requesting the referenced calculation files which documented the actual analyses cited on the SCEW sheets.

Since the SER Response submittal of November 1981, all previous analyses for safety-related equipment have been consolidated to provide easily traceable references for the equipment qualification Central File. These consolidated calculation files have been upgraded to provide the best available information for establishing documentation for equipment qualification, and provide the basis for the responses to deficiencies cited in the TER. These consolidated calculation files are available for inspection upon request.

Page A-6

Generic Item Number 3

The aging life established is often greater than the maintenance interval recommended by the manufacturer or by good engineering practice.

Response to Generic Item Number 3

The qualified life for thermal aging was derived on the basis of materials analysis, to predict how long a component can survive when exposed to normal plant conditions and postulated accident environments. The qualified life for thermal aging is not used to supersede or extend either the manufacturer's recommended service intervals or maintenance intervals which result from operating experience under Monticello's surveillance and maintenance program. Rather, the qualified life for thermal aging provides input to the surveillance and maintenance program by identifying potential weak links in the construction of safety-related electrical equipment.

L , I

Director of NRR Monticello Nuclear Generating Plant April 12, 1983

Attachment B

Page B-1

AVCO

Common Deficiencies and Responses

VENDOR: Automatic Valve Corporation EQUIPMENT DESCRIPTION: Solenoid Valves APPLICABLE EQUIPMENT ITEM NUMBERS: 34, 36

The FRC reviewed EDS calculation No. 0910-001-02, "AVCO Solenoid Valves," dated 10 Nov 81. This calculation file has been superseded by EDS calculation file No. 0910-106-EQ-03, in which the AVCO solenoid valves have been completely re-evaluated. The new file was not available to the FRC at the time of the audit and addresses the deficiencies listed by the FRC as shown below.

COMMON DEFICIENCIES AND RESPONSES:

1. "Adequate similarity between the installed and tested solenoid valves was not established."

RESPONSE:

Verified conversations with the manufacturer indicate that the C-4988-15 manifold assembly and solenoids have not been tested. However, the manufacturer also indicated test data are available for a similar assembly, the B-4988-3. The following specific differences exist between these two equipment items as identified by the manufacturer and component drawings:

a. The 4988-15 utilized potted class H solenoid coils, whereas the 4988-3 utilized non-potted class H solenoid coils.

This difference affects qualification in a positive sense, in that the potted coils are more resistant to steam/moisture intrusion. Thus, they are less prone to failure during an accident.

b. The 4988-15 incorporates one non-compensated plunger, whereas the 4988-3 has none.

The use of a non-compensated plunger on the C4988-15 is considered insignificant to the applicability of this Test Report because: 1) the compensated plunger merely has a metallic spring which helps return the plunger to the normal position; 2) the function of the unit remains the same; 3) there is no change in the materials of construction.

Page B-2

"Aging Degradation Inadequately Evaluated." "Criteria Regarding Aging Simulation Not Satisfied."

RESPONSE TO ITEMS 2 AND 3:

See Generic Response Item 1 in Attachment A.

The qualified life for this equipment is based on of a vendor-supplied materials list. The limiting aging is the silicone varnish. This material has thermal aging greater than 40 years for installed

materials analysis material for thermal a qualified life for locations.

4. "Criteria Regarding Steam Exposure Inadequate."

RESPONSE:

The cited test report, Rockwell Test Report 2792-03-02 (TER Reference 806), documents the results of LOCA environmental testing. The test procedure used saturated steam to maintain chamber temperature and to produce the thermal transients. This is considered adequate to ensure operability in a steam environment during postulated accidents.

2. 3.

Page B-3

BARKSDALE


VENDOR: Barksdale EQUIPMENT DESCRIPTION: Pressure Switch APPLICABLE EQUIPMENT ITEM NUMBERS: 51, 52, 53, 54, 55, 56, 57

The FRC reviewed EDS calculation No. 0910-001-010, "Barksdale Pressure Switch," and EDS calculation No. 0910-001-SER-5, "SER Response: Barksdale Pressure Switch". Barksdale Pressure Switches have been re-evaluated in EDS calculation No. 0910-106-EQ-25. This new file was not available to the FRC at the time of the audit and addresses the deficiencies listed by the FRC as shown below.


1. "An environmental qualification accident test profile does not constitute an accelerated thermal aging test program."

RESPONSE:


The qualified life for this equipment was re-evaluated by materials analysis of vendor-supplied materials list.

2. "The estimated qualified life of 15.85 hours is significantly less than the desired 40-year life, and the 6-hour exposure to 212oF is considerably less time than required to develop an adequate estimate of thermal aging life."

RESPONSE:

EDS calculation No. 0910-106-EQ-25 has extended the qualified thermal life of the Barksdale Pressure Switches to greater than 40 years as determined by material analysis. This qualified life is based on data predicting the qualified life of the limiting material, Buna-N. (See also Generic Response Item I in Attachment A.)

Page B-4

3. "Criteria Regarding Peak Pressure Inadequate."

RESPONSE:

The peak pressure specifications for each equipment group are identified below:

Item #51: This item is required only to mitigate the consequences of a LOCA, and is not required for a HELB. A LOCA does not affect the temperature, pressure, and relative humidity conditions at this location.

Item #52: 1.8 psig

Item #53: 0.6 psig

Items #54, 55, 56: 0.5 psig

Item #57: 0.1 psig

Pressure qualification for these equipment items is based on a review of the materials of construction and equipment design in comparison to the magnitude and duration of the pressure transient.

Pressure transients of up to 0.6 psig fall into the range of atmospheric fluctuations. Although the pressure change with respect to time is more rapid than for atmospheric changes, these excursions do not pose a credible failure mode to these pressure switches. Further, pressure transients of up to 1.8 psig do not pose a significant threat to the operability of this model of pressure switch. These devices are small (4"x7"x2 1/2"), constructed of heavy gage aluminum sealed with a neoprene gasket, and built to NEMA 4 specifications. This rugged construction is considered sufficient to withstand 1.8 psig on the basis of partial testing to .25 psig and its NEMA 4 rating. The NEMA 4 construction requires equipment to withstand 65 gallons per minute spray delivered from a 1" nozzle from a distance between 10 and 12 feet. This requirement has been equated to a static pressure of 4.7 psig using the Bernouli equation. Thus, operation during a pressure excursion of 1.8 psig is reasonably assured.

For Item #57, the thermal-hydraulic environment has been re-evaluated. The worst case pressure transient is now 0.1 psig, based on a feedwater line break in the Turbine Building at Elevation 931'. This peak is adequately enveloped by the testing of Test Procedure 9993, 6/23/75 (FRC TER reference PGR #3422).

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Page B-5

BARTON


VENDOR: Barton EQUIPMENT DESCRIPTION: D/P Indicating, Pressure, Flow, and D/P Switches APPLICABLE EQUIPMENT ITEM NUMBERS: 58, 71, 76, 78

The FRC reviewed EDS calculation No. 0910-001-011, Rev. 2, dated 12 Nov 81. This Barton pressure switch has been re-evaluated in EDS calculation No. 0910-106-EQ-06. This new file was not available to the FRC at the time of the audit and addresses the deficiencies listed by the FRC as shown below.


1. Documented evidence of qualification inadequate. FRC concluded "...that model 288 (289) has not been analyzed or tested and therefore lacks qualification documentation.

RESPONSE:

General Electric Company supplied environmental qualification records for General Electric-supplied NSSS safety-related electrical equipment for NSP-Monticello in answer to purchase order no. MQ05987. This document, dated November 1981, is further identified as DRF #AOO-1084, Index #34, DV 145C3008, NAME: Barton 289 Pressure Switch, and by EDS DI #0910-106-003, and will be identified here as DV145C3008. This Barton report details the testing of Barton models 288A and 289A pressure switches to simultaneous temperature, pressure, and relative humidity. Since the tested units are virtually identical to the earlier 288 and 289 switches, the test report also applies to these units. The difference between the old switches and the new switches are documented in verified conversations with the vendor. The only difference between them is that the newer, 'A', switches have a modified set point locking clip. This difference is insignificant to environmental qualification, because this part is made from metal.

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Page B-6

2. Aging degradation inadequately evaluated. 3. Qualified life or replacement schedule has not been established.



The qualified life has been evaluated in EDS calculation No. 0910-106-EQ-06 to be greater than 40 years as determined by material analysis. This qualified life is based on data predicting the life of the limiting material, Buna-N.

4. Criteria regarding peak pressure inadequate.

RESPONSE:

The most severe pressure profile peaks at 16.5 psia 42 seconds after accident initiation and returns to atmospheric within two minutes. The Barton pressure switch is re-evaluated in EDS calculation No. 0910-106-EQ-06, and qualified to greater than 16.5 psia by calculating an equivalent pressure for the NEMA 4 "hosedown" test. A stress analysis of the glass cover plate was performed to ensure the external pressure boundary remains intact. Therefore the pressure parameter is satisfied.

5. Required profile inadequately enveloped.

RESPONSE:

The operating time parameter is satisfied by testing for this equipment at the specified locations. The operating times for this equipment is 10 minutes. However, to ensure margin, one-hour operating times were demonstrated. The one-hour operating time is enveloped by testing at 212OF for six hours, with the following two exceptions:

a. FRC item 58; required operating time 10 minutes.

The peak accident temperature is 213 0F, 10F above the test temperature. Within 150 seconds, the accident temperature drops below the test temperature. This 1oF difference between the test temperature and accident temperature is considered insignificant for the following three reasons:

Page B-7

1) the test occurred for a period of 16 hours while the required operating time is 10 minutes;

2) the extra 10F for 100 seconds will cause an insignificant increase in thermal degradation; and

3) the thermal capacitance of the switch case would tend to mitigate the effects of the thermal spike seen by switch internals.

b. FRC item 71, required operating time 8 hours.

The required operating time of 8 hours is more than the test time of 6 hours. This is considered insignificant for the following reasons:

* The peak accident temperature of 127 0F for this equipment is well below the tested level of 212 0F. An Arrhenius calculation in EDS calculation No. 0910-106-EQ-06 demonstrates that the thermal aging due to testing is greater than the thermal aging due to the accident.

Enclosure 4 to IE Bulletin 79-01B, the DOR Guidelines, in paragraph 6.2.1 specifically allows extrapolation of test data to envelop the specified operating time, using the Arrhenius method.

6. Criteria regarding steam exposure is inadequate.

RESPONSE:

Barton Pressure Switches were tested in the cited test report (ITT Barton Test Report 9999.1712.2) to conditions which included 100% relative humidity. For the purpose of this evaluation, consideration of the temperature and 100% relative humidity parameters, as well as the relatively mild accident environments, satisfies the steam exposure criteria.

Page B-8

FENWAL


VENDOR: Fenwal Electric Co. EQUIPMENT DESCRIPTION: Fenwal Temperature Switch APPLICABLE EQUIPMENT ITEM NUMBERS: 85, 86, 87, 88

The FRC reviewed EDS calculation No. 0910-001-EER-lO, "EER Response: Fenwal Temperature Switch", dated 10 Oct 81. This calculation file has been superseded by EDS calculation No. 0910-106-EQ-09, dated 1-25-83, in which the Fenwal Temperature Switches have been re-evaluated. This new file was not available to the FRC at the time of the audit and addresses the deficiencies listed by the FRC as shown below.


1. Similarity has not been established between the test specimen and the installed models.

RESPONSE:

With respect to Fenwal Test Report No. 6350 (TER Reference 824):

Three temperature switches were tested, all model No. 17023-6. Documented conversations with the vendor indicate that the 17023-6 is a normally open temperature switch which closes on rising temperature, and the 17002-40 is a normally closed temperature switch which opens on rising temperature. Both temperature switches have identical non-metallic parts. Therefore, the test report referenced above is applicable to both model Nos. 17023-6 and 17002-40.

2. Steam exposure inadequate. More specifically, immersion testing provides some confidence that the unit is sealed and will function, however, the unit was not qualification tested under HELB conditions.

RESPONSE:

The Fenwal Engineering Data Report No. 6350, "Qualification of Fenwal 17023-6 units to General Electric Drawing No. 145C3004" (TER Reference 824), states that each switch was subjected to immersion in a wellagitated water bath maintained at 2001F and 10 psig. The switches functioned properly throughout this test and without any anomalies.

Page B-9

This test closely approximates steam exposure and provides reasonable assurance that these temperature switches will properly function during postulated HELB accident conditions which include 100% relative humidity.

3. Criteria regarding radiation inadequate; specifically, the tefloncoated leadwires are limited to a radiation exposure of 1.5x10 4 Rads or less.

RESPONSE:

The following paragraph is excerpted from EDS Calculation File No. 0910-106-EQ-09 and explains the radiation analysis done on the Fenwal Temperature Switch:

"The following non-metallic components have been identified by the vendor as materials of construction for these Fenwal Temperature Switches:

Material Radiation Threshold

MICA Insulation 1 x10 8 Rads TAGT

Teflon 3.7 x 104 Rads (25% Damage) Asbestos 9.8 x 107 Rads Glass (Fiberglass) 1 x 1011 Rads Teflon 3.7 x 104 Rads (25% Damage)

Glass Bead 1 x 1011 Rads Silicone Varnish 1 x 108 Rads Graphite 5 x 107 Rads

Because the TAGT is a layered insulation, each component will be evaluated individually. Since Teflon is on either side of the other two insulating materials, it is conservative to neglect its contribution to the electrical insulation of the leadwires:

The failure modes for Teflon include loss of mechanical strength coupled with a drop in volume resistivity; however, the mechanical strength and insulating properties of the Fiberglass and Asbestos are more than sufficient to compensate for this. The loss of mechanical strength would tend to cause the outer layer of Teflon to separate; the Fiberglass and Asbestos would keep the inner layer of Teflon intact so that any retention of insulating ability by the Teflon is a conservative addition to the overall insu ating properties of the

Page B-10

other materials. Therefore, Teflon is not essential to the operation of the switches and will not be considered in the evaluation.

The vendor indicates that the Silicone Varnish is used in minute quantities and is not significant to the operability of these switches. Graphite is used as a thread lubricant during the assembly of the switches and is not required for the switch to perform its safety function.

Therefore, the limiting material is Asbestos in the TAGT insulation, with a radiation damage threshold of 9.8 x 107 Rads. This exceeds all specified required doses, therefore, the radiation parameter is satisfied."

4. Aging degradation evaluated inadequately. 5. Qualified life or replacement schedule not established.



The qualified life for this equipment is based on materials analysis of a vendor-supplied material list. This equipment is qualified for a thermal life greater than 40 years for installed locations.

Page B-11

GENERAL ELECTRIC MOTORS


VENDOR: GE EQUIPMENT DESCRIPTION: Motors APPLICABLE EQUIPMENT ITEM NUMBERS: 99, 100, 101

The FRC reviewed EDS calculation No. 0910-001-041, Revision 3, "General Electric Motors." The GE motors have been re-evaluated in EDS calculation No. 0910-106-GEMOT-12, dated 5 Jan 83. This new file was not available to the FRC at the time of the audit and addresses the deficiencies listed by the FRC as shown below.


1. "Aging degradation evaluated inadequately." 2. "Qualifed life not established."


EDS calculation No. 0910-106-GEMOT-12 has extended the qualified thermal aging life of the GE motors to greater than 40 years, as determined by material analysis. Lubricating oil should be changed at scheduled intervals according to manufacturer's procedures, but not to exceed 8.5 years. See also Generic Responses, Items 1 and 3, in Attachment A.

Page B-12

LIMITOROUE


VENDOR: Limitorque Corporation EQUIPMENT DESCRIPTION: Motorized Valve Actuators APPLICABLE EQUIPMENT ITEM NUMBERS: 1, 2, 3, 5, 6, 7, 32, 148, 149, 150

FRC reviewed EDS calculation file No. 0910-001-224 Revision 4, "Limitorque Valve Motor Operators." (TER PSR No. 43) This calculation file has been superseded by EDS calculation No. 0910-106-EQ-14. This file was not available to FRC at the time of the audit and addresses the deficiencies listed in the TER.


1. "Documented Evidence of Qualification Inadequate."

RESPONSE:


2. "Adequate Similarity Between Equipment and Test Specimen Not Established."

RESPONSE:

The calculation file reviewed for the TER was completed before all of the applicable information had been received from Limitorque. Subsequent analyses of each valve operator was based on the application of test reports as prescribed by Limitorque. The applicability of specific test reports was decided by Limitorque after reviewing specific data concerning each operator. The information on which the decision of test report applicability was made included the Limitorque Order No. and the Actuator Serial No. From this information, Limitorque provided the Motor Serial No., Motor Type and Insulation Class, the applicable test report, and specific qualification concerns including the presence of brakes.

The motor manufacturers (Reliance Electric for AC Motors and Porter Electric for DC Motors) and brake manufacturer (where applicable)

Page B-13

were then contacted for detailed information concerning materials of construction for each motor. This information was used to perform detailed aging analyses to establish a qualified life for thermal aging or maintenance intervals and requirements to extend the life to 40 years. (See Items 3 and 4 below.)

3. "Aging Degradation Evaluated Inadequately." 4. "Qualified Life or Replacement Schedule Not Established."


See Generic Item 1 in Attachment A.

The qualified life/replacement schedule is being re-evaluated by materials analysis of the vendor-supplied materials lists.

Page B-14

MERCOID


VENDOR: Mercoid EQUIPMENT DESCRIPTION: Mercoid Pressure Switches (HPCI Room) APPLICABLE EQUIPMENT ITEM NUMBERS: 62, 63

FRC reviewed EDS calculation No. 0910-001-028, "Mercoid Pressure Switches" (TER Reference No. 54). This calculation file has been superseded by EDS calculation No. 0910-106-EQ-28, which was not available to the FRC at the time of the audit. The new file addresses the deficiencies listed in the TER.


1. "Aging degradation inadequately evaluated."

RESPONSE:

The TER identified the source of this deficiency as conflicting statements in the thermal aging analysis. EDS calculation No. 0910-106-EQ-28 resolves this deficiency by clearly indicating which notes are applicable to each of the non-metallic materials.

2. "Criteria Regarding Humidity Not Satisfied." 3. "Criteria Regarding Pressure Not Satisfied."


Plant-specific thermal-hydraulic analyses have been performed to establish the temperature and pressure response of plant spaces containing safety-related equipment items in response to a variety of accident scenarios. The most severe accident for the HPCI Room is an HPCI line break in the HPCI Room; however, this accident would render the HPCI system inoperable and obviate the operation of this equipment. The most severe response for which the HPCI system must operate is presented in Figure B-6 of the November SER submittal (TER Figure A-7). As can be seen from this response profile, the HPCI Room is subjected to only a 20F temperature transient and a 0.4 psig pressure transient. This response, based on worst case initial conditions, is

Page B-15

clearly non-harsh. In addition, the HPCI room is isolated from the postulated pipe break which produces this transient by a tank room. This intermediate space will prevent the relative humidity in the HPCI Room from being significantly affected. Thus, the impact of all accident conditions except radiation can be neglected, as they are not significantly more severe than normal.

Page B-16

ROSEMOUNT


VENDOR: Rosemount EQUIPMENT DESCRIPTION: Differential Pressure Transmitter APPLICABLE EQUIPMENT ITEM NUMBERS: 93, 147

FRC reviewed EDS calculation No. 0910-001-052, "Rosemount Pressure Transmitters" (TER Reference No. 54). This calculation file has been superseded by EDS calculation No. 0910-106-EQ-24, which was not available to FRC at the time of the audit. The new file addresses the deficiencies listed in the TER.



RESPONSE:


2. "Aging Degradation Evaluated Inadequately." 3. ."Qualified Life or Replacement Schedule Not Established."



The qualified life for this equipment.is based on materials analysis of vendor-supplied material list and test results from the Westinghouse Test Report (TER Reference 4716). The limiting material for thermal aging is an epoxy glass laminate circuit board. This material has a qualified life for thermal aging greater than 40 years for installed locations. See also Response to Items 5, 6, and 7 below.


RESPONSE:

The TER identified that the source of this deficiency is the absence of a cable connection interface. "Westinghouse Qualification Testing

Page B-17

of Rosemount Pressure Transmitters, Model 1153 Series A" (TER Reference 4716), identifies steam entry into the electronics as a chief cause of testing anomalies; therefore, the interface is critical for successful operation under DBE conditions.

RESPONSE:

Northern States Power Company will install a qualified seal assembly to ensure the conduit pipe threads are sealed.

"Criteria "Criteria "Criteria

Regarding Radiation Not Satisfied." Regarding Test Failures or Severe Anomalies Not Satisfied." Regarding Instrument Accuracy Not Satisfied."

RESPONSE TO ITEMS 5, 6, AND 7:

Each of these deficiencies results from the testing anomalies documented in the Westinghouse Test Report (TER Reference 4716). The testing documented in this report is used to evaluate thermal aging only. The Rosemount Test Report (TER Reference 1764) successfully tested three transmitters to conditions which more closely simulated postulated accident environments of installed locations, and is used to qualify this equipment for all other parameters.

The Westinghouse Test Report (TER Reference 4716) subjected test specimens to the conditions listed below; these conditions are contrasted with specified conditions at Monticello.

Parameter

Dose Rate

Total Integrated Dose

Temperature

Pressure

Chemical Spray

Monticello Value

5.5 x 105 Rads/hr. (conservative)

1 x 107 Rads

213.8 0F (max)

17.04 psia

None

Tested Value

1.5 x 106 Rads/hr

5 x 107 Rads/hr

370aF spike

Not specified (Psat=92 psia)

Boric Acid/Sodium Hydroxide, pH 9.25

5. 6. 7.

Page B-18

As can be seen, each test parameter is more severe than the requirements for that parameter at Monticello. As a result of the test severity, several anomalies/failures were induced during the testing. It is nevertheless acceptable and conservative to use this testing to evaluate the equipment for thermal aging. The failures (end-of-life condition) were not due to thermal degradation alone; but, by assuming that it was and extrapolating the test data, a conservative thermal aging life can be established. See also response to items 2 and 3 above.

The following evaluations of each test sample were provided to justify the use of the Westinghouse Test Report (TER Reference 4716) to evaluate these transmitters for thermal aging:

a. Model 1153 GA9 (AW-1):

Largest error occurred during radiation exposure. This error was 50.2% of the span or 13.5% of the upper range limit. During steam/ chemical testing, maximum error was 13% of span. The transmitter performed throughout the test and was within 4.3% of span after completion.

The large error that occurred during radiation testing was attributed to the dose rate and not the total integrated dose. However, this error is not typical of 1153 Series A transmitters as shown by the performance of the other three transmitters in this test and the transmitters tested in the Rosemount Test.

b. Model 1153 GA9 (AW-2)

Worst case performance during radiation testing was 4.0% of span (4.0% of upper range limit). Output was erratic throughout the test procedure. In the 15th day of testing, the output went to a low of 2.9 mA value.

The constant 2.19 mA output was attributed to a short between the primary and feedback windings of the oscillator transformer. Erratic output was attributed to a sensor defect. It was judged that this failure could appear during normal operation and was not a result of qualification testing. This failure has not been witnessed in previous testing.

c. Model 1153 DA5 (AX-1)

Worst case performance during radiation testing was 13.2% of span (8.7% of upper range limit). The worst occurred during steam/ chemical testing at 7.0% (4.3% of upper range limit). The output

Page B-19

signal went high after 8 days and returned to normal periodically until 12 days, when it remained high for the rest of the test.

The output errors were attributed to a defective output stage op-amp. The failure of this op-amp was due to a defective or degraded seal which allowed contaminants to enter the die cavity. This is considered a random failure.

d. Model 1153 DA5 (AX-2)

Worst case performance during radiation testing was -15.9% of span (-5.6% of upper range limit) during the steam/chemical testing; the error was 28% of span .(9.9% of upper range limit). However, the Conax fitting had leaked which allowed the caustic spray solution to enter the housing. After cleaning, the transmitter operated for 3 days before high output failure occurred.

The cause of the high output failure was again attributed to a short between the primary and feedback winding of the transformer. The 28% error was attributed to failure of the terminal block positive terminal signal. This was due to a Conax fitting leak during the steam test. After replacing the faulty electronics, the transmitter was retested at 330 0F. The error was 13% of span (4.6% of upper range limit). This indicates that failure was, in fact, due to caustic spray entering the transmitter.

In summary, the Model 1153 DA5 (AX-1) was used to determine qualification for the following reasons:

Model 1153 DA5 (AX-2) should not be included because failure was due to a Conax fitting rather than the transmitter itself. The Conax fittings are evaluated separately.

Model 1153 DA9 (AW-1) performed throughout the test procedure. However, it is not conservative to use this as a representative of all 1153 Series A transmitters since the failure displayed by AX-1 is a possible failure mode for the installed transmitters.

Model 1153 GA9 (AW-2) failed in the 15th day of testing due to the same op-amp failure that occurred in AX-1. However, failure of this transmitter occurred after the AX-1 failure. For conservatism, the AX-1 results are used.

Page B-20

Although Model 1153 DA5 (AX-1) operated until the 12th day of testing, erratic output first occurred after the 8th day. The aging evaluation is based only on the first 8 days of testing. This conservatively predicts the thermal aging qualification of this transmitter. See Item No. 2 above for a further description of aging methodology.

Page B-21

STATIC-O-RING


VENDOR: Static-O-Ring EQUIPMENT: Pressure Switch APPLICABLE EQUIPMENT ITEM NUMBERS: 65, 66, 67, 68, 69, 70

The FCR reviewed EDS calculation No. 0910-001-SER-32, "Static-O-Ring Pressure Switch," 10 Nov 81. The Static-O-Ring pressure switch has been re-evaluated in EDS calculation No. 0910-106-EQ-26, Rev. 1, February 3, 1983. This new file was not available to the FRC at the time of the audit and addresses the deficiencies listed by the FRC as shown below.


1. Inadequate evidence of qualification documentation. "The licensee did not furnish supporting documentation for the aging and radiation analysis to permit an independent review."

RESPONSE:


EDS calculation No. 0910-106-EQ-26, Rev. 1, re-evaluates the Static-0Ring pressure switches. It is considered that this new file will meet the FRC concerns of providing a comprehensive analysis of the Static0-Ring pressure switches that could be verified by an independent reviewer.

2. "Qualified life or replacement schedule not established."

RESPONSE:


The qualified life for these items is based on materials analysis of a vendor-supplied materials list. The limiting material for thermal aging is an epoxy. Thermal aging qualification for these pressure switches is therefore based on the thermal aging capabilities of the epoxy. This material has a qualified life for thermal aging greater than 40 years for the installed locations.

Page B-22

YARWAY


VENDOR: Yarway Corporation EQUIPMENT DESCRIPTION: Yarway Level Indicator APPLICABLE EQUIPMENT ITEM NUMBERS: 81, 83

The FRC reviewed EDS calculation No. 0910-001-038, Rev. 2, "Yarway Level Indicator/Transmitter Switch," dated 10 Nov 81. This calculation file has been superseded by EDS calculation No. 0910-106-EQ-18, dated 1-25-83, in which the Yarway Level Switches have been re-evaluated. This new file was not available to the FRC at the time of the audit and addresses the deficiencies listed by the FRC as shown below.


1. "Inadequate similarity established between the tested and installed equipment."

RESPONSE:

The following paragraph was taken from EDS calculation No. 0910-106-EQ-28:

The above test [7] reports that a Yarway 4418TC level indicator/transmitter was subjected to simultaneous temperature/humidity testing (see page 16 of this file).

* Refs. [8, 15] indicate that the above test report is applicable to model number 4418C being evaluated here.

* The 4418EC is a 4418C module with an additional transmitter/ armature assembly, for remote indication.

* The 4418C indicators currently installed at Monticello employ mercury switches, rather than the micro switches tested in [7] or the available Acro switches. (The mercury switches are addressed by analysis.)

Page B-23

2. "Criteria for peak pressure inadequate. More specifically, Licensee does not provide evidence that the unit will function under HELB steam/pressure conditions since no pressure testing was done."

RESPONSE:

The postulated accident pressure spike of 0.6 psig falls into the range of normal atmosphere fluctuations. Although the pressure change with respect to time is more rapid than for atmospheric changes, the excursions postulated do not pose a credible failure mode to these level switches.

3. "Radiation qualification has not been established."

RESPONSE:

The following information was taken from EDS calculation No. 0910-106-EQ-28.

Non-metallic components of the 4418C indicator were identified by the vendor and a literature search was performed to determine the radiation damage threshold of each. The limiting material (Neoprene) has a radiation damage threshold of 8x10 5 Rads, which exceeds the maximum required dose, 2.4x10 5, including margin. Therefore, the radiation parameter is satisfied. The only non-metallic part in the mercury switch is the glass capsule. Radiation damage threshold for glass is presented below.

The following non-metallic materials necessary for operation of the Yarway 4418C indicator were identified in Ref. [9]. Each is listed with its radiation damage threshold. .

--I

Page B-24

Material

Dacron Polyester

Ethylene Propylene

Buna 'N'

Neoprene

DK149 Cork/* Rubber Compound

Insulated Vinyl Tubing

Glass

* Ref. [10] indicates Neoprene is assumed evaluation.

Radiation Damage Threshold (Rads)/ Failure Mode

4.4 x 106/tensile/elongated

lx 106/compression set

I x 106/compression set

8 x 105/compression set, loss of flexibility

8 x 105/compression set, loss of flexibility

1 x 107/ASTM D1672

1 x 1011/none

that DK-149 is a Neoprene/cork compound. to be limiting for the purpose of this

5. An environmental qualification test profile does not constitute an accelerated thermal aging test program.

RESPONSE:


6. Aging Degradation Inadequately Evaluated. 7. Qualified Life or Replacement Schedule Not Established. 8. Program to Identify Aging Degradation Inadequate. 9. Criteria Regarding Aging Simulation Inadequate.

RESPONSE TO ITEMS 6, 7, 8, AND 9

Thermal aging qualification for these level indicators is established by evaluating the susceptibility of non-metallic component materials to deterioration from thermal aging using the Arrhenius Method. This

Page B-25

qualified life is based on a vendor-supplied materials list, materials test data, and specified normal and accident temperatures. The limiting material for thermal aging was identified as cork/rubber, used as a gasket material. These level switches are qualified to perform their required safety function during postulated accident conditions after exposure to specified normal conditions for 9.5 years. To maintain qualification of the equipment the cork/rubber gaskets must be replaced at intervals of less than 9.5 years. See also Generic Response for item 7 in Attachment A.

REFERENCES:

[7] Lockheed Test Report #5628-3509, "Yarway Remote Level Indicator M/N 4418TC, dated 3/27/79, EDS DI No. 0910-001-068.

[8] EDS Materials Data Base, EDS DI No. 0910-106-009.

[9] Materials List for Yarway 4400-C Level Indicator received from W.A. Huss (Yarway), dated 9/8/80, EDS DI No. 0910-001-110.

[10] EDS Letter No. 0910-106-018 to John Yakubowski (Armstrong) from L.D. McHugh (EDS) dated 8/16/82, EDS DI No. 0910-106-032.

[15] EDS Letter No. 0910-106-017 with ROC from L.D. McHugh (EDS) to W.A. Huss (Yarway), dated 8/13/82, EDS DI No. 0910-106-070.

[17] Yarway Letter from W.A. Huss (Yarway) to L.D. McHugh (EDS) with enclosures, 19 October 1982, EDS DI No. 0910-106-065.

Director of NRR April 12, 1983

Monticello Nuclear Generating Plant

Attachment C

Page C-1

ITEM NUMBER: DESCRIPTION: Limitorque Motorized Valve Operator;

with Class RH, AC motorsLOCATION: Containment PLANT ID NUMBERS: MO 2-53 (A,B) and 2-54 (A,B) NRC QUALIFICATION CATEGORY: II.a

SMB, SB series

ITEMIZED DEFICIENCIES AND RESPONSES:

.1. Documented Evidence of Qualification Inadequate."

RESPONSE:

See Limitorque Common Deficiency #1 and Response in Attachment B.


RESPONSE:


3. 4.

"Aging Degradation Evaluated Inadequately." "Qualified Life or Replacement Schedule Not Established."


See Limitorque Common Deficiencies #3 and #4 with Response in Attachment B.

EQUIPMENT EQUIPMENT

Page C-2

EQUIPMENT ITEM NUMBER: 2 EQUIPMENT DESCRIPTION: Limitorque Motorized Valve Operator; SMB, SB series

with Class H, AC motors LOCATION: Containment PLANT ID NUMBERS: MO 2034, 2379, 2027, 2029, 2373, 2075 NRC QUALIFICATION CATEGORY: II.a


T. Documented Evidence of Qualification Inadequate."

RESPONSE:



RESPONSE:





Page C-3


with Class B, DC motors; Motor Brake Installed LOCATION: Steam Chase PLANT ID NUMBERS: MO 2068 NRC QUALIFICATION CATEGORY: II.a


Northern States Power Company will replace this equipment with a fully qualified model.

Page C-4


with Class H, DC motors LOCATION: Steam Chase PLANT ID NUMBERS: MO 2076 (MO 2035 and 2374 unit motors have been replaced

with qualified Class RH motors) NRC QUALIFICATION CATEGORY: II.a


1. Documented Evidence of Qualification Inadequate."

RESPONSE:



RESPONSE:



RESPONSE TO ITEMS 3 AND 4: '


Page C-5


with Class H, DC motors LOCATION: Reactor Building, Elev. 975' PLANT ID NUMBERS: MO 20398, 2026 NRC QUALIFICATION CATEGORY: II.a


1. Documented Evidence of Qualification Inadequate."

RESPONSE:

See Limitorque Common Deficiency #1 and Response in-Attachment B.


RESPONSE:





Page C-6

EQUIPMENT ITEM NUMBER: 8 EQUIPMENT DESCRIPTION: Limitorque Motorized Valve Operator;

with Class H, AC motors; Motor Brake LOCATION: Reactor Building Elev. 948' PLANT ID NUMBERS: MO 1751, 1752, 1753, 1754 NRC QUALIFICATION CATEGORY: II.a

SMB, SB series installed



Page C-7

EQUIPMENT ITEM NUMBER: 30 EQUIPMENT DESCRIPTION: ASCO Solenoid Valve Model WPLBX831636 LOCATION: Reactor Building, Elev. 950'O" NE PLANT ID NUMBERS: SV3-140A,B NRC QUALIFICATION CATEGORY: II.a

The FRC reviewed EDS calculation No. 0910-001-009, Rev. 2, "ASCO Solenoid Valves," dated 10 Nov. 81. This problem file has been superseded by EDS calculation No. 0910-106-EQ-02, in which the ASCO solenoid valves have been completely reevaluated. The new file was not available to the FRC at the time of the audit and addresses the deficiencies listed by the FRC as shown below.


1. "Aging Degradation Inadequately Evaluated." 2. "The License Has Not Established a Qualified Life Estimate for This

Equipment." 3. "Program To Identify Aging Degradation not Established."

RESPONSE:

See Generic Response for Item I in Attachment A.

EDS calculation No. 0910-106-EQ-02 has shown that the thermal life of the equipment is greater than 40 years, using Arrhenius methodology.

Page C-8

EQUIPMENT ITEM NUMBER: 31 EQUIPMENT DESCRIPTION: ASCO LOCATION: Reactor Building, PLANT ID NUMBERS: SV3-3TA,B, NRC QUALIFICATION CATEGORY:

Solenoid Valve Model HVA 90441TA Elev. 935' W SV3-13-TT7, SV3-T3-118 II.a

The correct model numbers for these items are as follows:

a. SV3-31A,B - ASCO, Model HVA 904411A

These solenoid valves have been replaced with fully qualified models.

b. SV3-13-117 and SV3-13-118 - ASCO, Model HVA 90-405


1. 2. 3.

"Aging Degradation Not Evaluated Adequately" "Qualified Life or Replacement Schedule Not Established" "Criteria Regarding Aging Simulation Not Satisfied"


See Generic Response for Item 1 in Attachment A.

The qualified life for thermal aging for these solenoid valves was evaluated from vendor-supplied materials lists. Since the calculated thermal aging life is less than 40 years, these items have been placed in a surveillance and maintenance program for periodic replacement of potentially degraded sub-components.

Page C-9

EQUIPMENT ITEM NUMBER: 34 EQUIPMENT DESCRIPTION: Automatic Valve Corp. Solenoid Valve Model C-4988-15 LOCATION: Containment PLANT ID NUMBERS: SV 2-80 (A-D) NRC QUALIFICATION CATEGORY: II.a


1. "Adequate Similarity Between Equipment and Test Specimen Not Establ i shed."

RESPONSE:

See AVCO Common Deficiency #1 and Response in Attachment B.

2. "Aging Degradation Inadequately Evaluated." 3. "Criteria Regarding Aging Simulation Not Satisfied."


See AVCO Common Deficiencies #2 and #3 with Response in Attachment B.

4. "Criteria Regarding Peak Temperature Inadequate." 5. "Required Profile Inadequately Enveloped."


The applicable test report, Rockwell Report 2792-03-03 (TER Reference 806), documents testing of assemblies similar to those installed. (See item #1 above.) This testing subjects the specimens to the following conditions:

3.3 hours at 340*F 7.2 hours at 310*F 3.8 hours at 250*F

Since these items are required for only 5 minutes at a peak temperature of 3090F, the test adequately envelopes postulated conditions.

Page C-10


See AVCO Common Deficiency #4 with Response in Attachment B.

7. "Criteria Regarding Chemical Spray Not Satisfied."

RESPONSE:

Correspondence with the equipment manufacturer indicates that the solenoid assemblies are potted to NEMA 4 specifications. Construction to NEMA 4 specifications assures operability in a spray (water) environment. Containment spray at Monticello consists of only demineralized water; therefore this parameter is considered satisfied.

Page C-11

EQUIPMENT ITEM NUMBER: 35 EQUIPMENT DESCRIPTION: Automatic Valve Corp. Solenoid Valve Model C-5450 LOCATION: Containment PLANT ID NUMBERS: SV 2-71 A,B,D; SV 2-71 (A-H) NRC QUALIFICATION CATEGORY: II.a




See AVCO Common Deficiencies #2 and #3 with Response in Attachment B.

3. "Criteria Regarding Peak Temperature Inadequate." 4. "Required Profile Inadequately Enveloped."


The test temperature profile (Tmax 340"F) of GE PED memo 126-62 envelops the accident profile (Tmax 309*F) at all times except between 6 and 10 hours, when the test specimen temperature was 250*F and accident conditions require 2810F. This discrepancy is insignificant because, whereas the accident profile terminates at 10 hrs, the test continues at 250*F to 24 hours, at which time the entire 24-hour sequence is repeated four more times. The accumulated time at 340OF (15 hrs) for the test alone exceeds the entire required operating time for the valve. Additional conservatism comes from the thermal transients endured by the test specimens. That is, subjecting the test specimens to cycles of 3 hours at 340*F, 3 hours at 320 0F, and 18 hours at 250'F five separate times is more severe than 15 hours at 340*F followed by 15 hours at 340*F followed by 90 hours at 2500 F. This added severity results from internal thermal stresses induced by the temperature transients. Therefore, the temperature conditions required by the accident profile are satisfied by the test.


RESPONSE:

GE PED Memo 126-62 describes that these solenoid valves were exposed to the temperature profile described above. Temperature was controlled

Page C-12

by two electric strip heaters on the side of the vessel and a heater at the bottom of the vessel. The bottom of the vessel was also covered with three inches of water. By maintaining water at the bottom of the chamber, and controlling temperature with these heaters, saturated conditions are approximated during the 90 hours at 250 0F. These conditions reasonably assure operability in a steam environment.

6. "Criteria Regarding Spray Not Satisfied."

RESPONSE:

Although the aforementioned testing did not include spray conditions, these solenoid valves are built to NEMA 4 specifications. Construction to NEMA 4 specifications assures operability in a spray (water) environment. Containment Spray at Monticello consists of only demineralized water, so this parameter is satisfied.

7. "Criteria Regarding Functional Testing Not Satisfied."

RESPONSE:

The applicable test report, GE PED Memo 126-62, documents a complete set of functional tests before, during, and after the environmental testing (800 total cycles). Therefore, this parameter is considered satisfied.

Page C-13

EQUIPMENT ITEM NUMBER: 36 EQUIPMENT DESCRIPTION: Automatic Valve Corp. Solenoid Valve Model C-4988-15 LOCATION: Steam Chase PLANT ID NUMBERS: SV 2-86 (A-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:




See AVCO Common Deficiencies #2 and 3 with Response in Attachment B.

4. "Criteria Regarding Steam Exposure Not Adequate."

RESPONSE:


Page C-14

EQUIPMENT ITEM NUMBER: 39 EQUIPMENT DESCRIPTION: Microswitch Limit Switch, Model OP-AR LOCATION: SGTS Room PLANT ID NUMBERS: Position Indication for A02944, A02978, AO2979 NRC QUALIFICATION CATEGORY: II.a

The FRC reviewed EDS calculation No. 0910-001-029, "Microswitch Limit Switch." This Microswitch has been re-evaluated in EDS calculation No. 0910-106-018. This new file was not available to the FRC at the time of the audit and addresses the deficiencies as shown below.


1. "Criteria Regarding Radiation Not Satisfied."

RESPONSE:

The re-evaluation of these Limit Switches has concluded that the previously referenced testing for the BZ-2R switches (TER Reference 4208) is sufficient to qualify them for opeation at installed locations. The BZ-2R switch mechanism is the active component in this equipment. It is mounted in a sealed aluminum housing and activated by a metal plunger assembly. The elastomeric components, all associated with the housing seals, do not affect the ability of the limit switch to perform its function, because there are no harsh thermalhydraulic conditions for the installed location. The only other nonmetallic component in the housing is the Fiberglass cloth insulation. This material has a damage threshold in excess of 8x10 7 Rads.

The BZ-2R mechanism has been tested to 1.2x10 7 Rads gamma (TER Ref 4208), which also exceeds the specified accident radiation dose. These limit switches are therefore qualified to operate at installed locations during postulated accident conditions.

Page C-15

EQUIPMENT ITEM NUMBER: 44 EQUIPMENT DESCRIPTION: NAMCO Limit Switch Model EA 170-34101 LOCATION: HPCI Room PLANT ID NUMBERS: HPCI Turbine Stop Valve Position NRC QUALIFICATION CATEGORY: II.a

The FRC reviewed EDS calculation No. 0910-001-EER-25, dated 72 Nov 81. The NAMCO limit switch has been re-evaluated in EDS calculation No. 0910-106-EQ-19. This new file was not available to the FRC at the time of the audit and addresses the deficiencies as shown below.



The materials list, threshold values, and data sources of radiation qualification were not identified in the submittal.

RESPONSE:

The non-metallic materials used in the construction of NAMCO EA-770-34101 have been identified in correspondence with the manufacturer. These materials are listed below along with their radiation threholds. The lowest radiation threshold for any of the non-metallic materials is 8.0x10 5 Rads gamma; thus, the radiation requirement of 2.5x105 Rads gamma is satisfied.

Material

Glass-Filled Polyester

Mineral-Filled ATkyd

GTass-Filled Polyester

Mineral-Filled Alkyd

Buna-N

Buna-N Coated Nylon

Glass-Filled Polyester

Polyvinyl Chloride

Formica (Pheonolic/Melamine)

Melamine

Radiation Threshold (Rads)

1.0 x 70 9

1.0 x TO7

1.0 x 10

1.0 x TO7

2.0 x 106

2.0 x 10 6/8.0 x 105

1.0 x 10

2.0 x 10 7

5.0 x 106/8.0 x 106

8.0 x 106

Ref.

23

25

23

25

23

23

23

23

23

23

Component

Contact Block

Contact Lever

0-Ring

Gaskets

Cover Insulator

Page C-16

REFERENCES:

[91 Letter R. H. Kantner (NAMCO) to M. Chibber (EDS) dated July 17, 1980. EDS DI # 0910-001-063.

[23] "A Review of Equipment Aging Theory and Technology," EPRI Report NP-1558, September 1980.

[25] "The Effects of Nuclear Radiation on Elastomeric and Plastic Components and Materials," Radiation Effects Information Center Report No. 21, EDS DI #0910-106-012.

Page C-17

EQUIPMENT ITEM NUMBER: 45 EQUIPMENT DESCRIPTION: NAMCO Limit Switch Model EA 740-50100 LOCATION: Containment PLANT ID NUMBERS: Position Indication for AO 2-80 (A-D) NRC QUALIFICATION CATEGORY: II.a



The Licensee has not identified any sealing tape was used for sealing the pipe thread. for steam exposure is inadequate because of connection method.

method, although Teflon Therefore, qualification lack of identification of

RESPONSE:

The Northern States Power Company will install a qualified seal assembly to ensure the pipe threads are sealed.

Page C-18

EQUIPMENT ITEM NUMBER: 46 EQUIPMENT DESCRIPTION: NAMCO Limit Switch Model EA 740-8000 LOCATION: Containment PLANT ID NUMBERS: Position Indication for CY-2790 NRC QUALIFICATION CATEGORY: II.a


Northern States Power Co. will replace this equipment with a fully qual if ied model.

Page C-19

EQUIPMENT ITEM NUMBER:. 49 EQUIPMENT DESCRIPTION: NAMCO Limit Switch Model SL5C3L LOCATION: Torus Compartment PLANT ID NUMBERS: Position Indication for CV-2384 NRC QUALIFICATION CATEGORY: II.a


Northern States Power Co. will replace this equipment with a fully qualified model.

EQUIPMENT ITEM NUMBER: 50 EQUIPMENT DESCRIPTION: NAM( LOCATION: Steam Tunnel PLANT ID NUMBERS: Position NRC QUALIFICATION CATEGORY:

CO Limit Switch Model SL382W

Indication for A02-86 (A-D) II.a



Page C-20

Page C-21

EQUIPMENT ITEM NUMBER: 51 EQUIPMENT DESCRIPTION: Barksdale Pressure Switch Model B2TAl2SS LOCATION: Reactor Bldg., West Side, Elev. 935'O" PLANT ID NUMBERS: PS 2-3-49 (A, B), PS 2-3-50 (A, B) NRC QUALIFICATION CATEGORY: II.a


1. "Aging Degradation Evaluated Inadequately."

RESPONSE:

See Barksdale Common Deficiencies #1 and #2 with Responses in Attachment B.


RESPONSE:

See Barksdale Common Deficiency #3 and Response in Attachment B.

Page C-22

EQUIPMENT ITEM NUMBER: EQUIPMENT DESCRIPTION: LOCATION: RHR Room PLANT ID NUMBERS: PS NRC QUALIFICATION CATE

52 Barksdale Pressure Switch Model B2TAT2SS

14-44 (A-D) GORY: II.a



RESPONSE:



RESPONSE:


Page C-23

EQUIPMENT ITEM NUMBER: 53 EQUIPMENT DESCRIPTION: Barksdale Pressure Switch Model B2TA12SS LOCATION: iReactor Bldg., West Side, Elev. 935'O" PLANT ID NUMBERS: PS 2-3-53 (A, B) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:


Page C-24

EQUIPMENT ITEM NUMBER: 54 EQUIPMENT DESCRIPTION: Barksdale Pressure Switch Model B2TA12SS LOCATION: Reactor Bldg., East Side, Elev. 935'O" PLANT ID NUMBERS: PS 23-68 (A-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:


Page C-25

EQUIPMENT ITEM NUMBER: 55 EQUIPMENT DESCRIPTION: Barksdale Pressure Switch Model 82TAT2SS LOCATION: Reactor Bldg, South Side, Elev. 962'6" PLANT ID NUMBERS: PS 2-3-51 (A,D) NRC QUALIFICATION CATEGORY: II.a


T. "Aging Degradation Evaluated Inadequately."

RESPONSE:



RESPONSE:


Page C-26

EQUIPMENT ITEM NUMBER: 56 EQUIPMENT DESCRIPTION: Barksdale Pressure Switch Model B2TAT2SS LOCATION: Reactor Bldg., South Side, Elev. 962'6" PLANT ID NUMBERS: PS 2-3-52A NRC QUALIFICATION CATEGORY: II.a



RESPONSE:

See Barksdale Common Deficiencies #1 and #2 with Responses in Attachment 8.


RESPONSE:

See Barksdale Common Deficiency #3 with Response in Attachment B.

Page C-27

EQUIPMENT ITEM NUMBER: 57 EQUIPMENT DESCRIPTION: Barksdale Pressure Switch Model B2TAT2SS LOCATION: Turbine Bldg.,.SE corner PLANT ID NUMBERS: PS 5-14 (A-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:


2. "Criteria Regarding Peak Temperature Inadequate."

RESPONSE:

The thermal-hydraulic environment for this equipment has been re-evaluated. The limiting accident for this location is a feedwater line break in the Turbine Building at Elevation 931'.

Both the peak temperature specified and reported are 212 0F, however, the accident remains at this peak for only a few seconds while the test remains at 212*F for 6 hours. This longer duration at the peak temperature provides adequate margin and reasonable assurance that this equipment will function during specified accident conditions.


RESPONSE:


Page C-28

EQUIPMENT ITEM NUMBER: 58 EQUIPMENT DESCRIPTION: Barton Pressure Switch Model 288 LOCATION: Reactor Bldg., South Side, Elev. 962'6" PLANT ID NUMBERS: PS 2-3-52B NRC QUALIFICATION CATEGORY: II.a


1. Documented evidence of qualification inadequate.

RESPONSE:

See Barton Common Deficiency #1 and Response in Attachment B.

Page C-29

EQUIPMENT ITEM NUMBER: 60 EQUIPMENT DESCRIPTION: Meletron Pressure Switch Model 3726 SS49A LOCATION: Reactor Bldg., East Side, Elev. 935'O" PLANT ID NUMBERS: PS-13-87 (A-D) NRC QUALIFICATION CATEGORY: II.a



Page C-30

EQUIPMENT ITEM NUMBER: 62 EQUIPMENT DESCRIPTION: Mercoid Pressure Switches LOCATION: HPCI Room PLANT ID NUMBERS: PS 23-84 NRC QUALIFICATION CATEGORY: II.a


1. "Aging Degradation Inadequately Evaluated."

RESPONSE:

See Mercoid Common Deficiency #1 and Response in Attachment B.

2. 3.

"Criteria Regarding Humidity Not Satisfied." "Criteria Rgarding Pressure Not Satisfied."


See Mercoid Common Deficiencies #2 and #3 and ment B.

Response in Attach-

4

Page C-32

EQUIPMENT ITEM NUMBER: 65 EQUIPMENT DESCRIPTION: Static-O-Ring Pressure Switch Model 12NAA4 LOCATION: Reactor Bldg., South Side, Elev. 962'6" PLANT ID NUMBERS: PS 10-101 (A-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:

See Static-O-Ring Common Deficiency #1 and Response in Attachment B.

2. Qualified life not established.

RESPONSE:


Page C-33

EQUIPMENT ITEM NUMBER: 66 EQUIPMENT DESCRIPTION: Static-O-Ring Pressure Switch Model 12NAA4 LOCATION: Reactor Bldg., South Side, ETev. 962'6" PLANT ID NUMBERS: PS 10-100 (A-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:


Page C-31

EQUIPMENT ITEM NUMBER: 63 EQUIPMENT DESCRIPTION: Mercoid Pressure Switches LOCATION: HPCI Room PLANT ID NUMBERS: PS 23-97 (A,B) NRC QUALIFICATION CATEGORY: II.a


1. "Aging Degradation Inadequately Evaluated."

RESPONSE:

See Mercoid Common Deficiency #1 and Response in Attachment B.

2. "Criteria Regarding Humidity Not Satisfied." 3. "Criteria Rgarding Pressure Not Satisfied."


See Mercoid Common Deficiencies #2 and #3 and Response in Attachment B.

I

Page C-34

EQUIPMENT ITEM NUMBER: 67 EQUIPMENT DESCRIPTION: Static-O-Ring Pressure Switch Model 12NAA4 LOCATION: Reactor Bldg., South Side, Elev. 962'6" PLANT ID NUMBERS: PS 10-119 (A-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:


Page C-35

EQUIPMENT ITEM NUMBER: 68 EQUIPMENT DESCRIPTION: Static-O-Ring Pressure Switch Model 5NAA3X LOCATION: RHR Room PLANT ID NUMBERS: PS 10-105 (A-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:


Page C-36

EQUIPMENT ITEM NUMBER: 69 EQUIPMENT DESCRIPTION: Static-O-Ring Pressure Switch Model 6NAA3 LOCATION: Reactor Bldg., West Side, Elev. 935'0" PLANT ID NUMBERS: PS 2-128 A,B NRC QUALIFICATION CATEGORY: II.a


T. Documented evidence of qualification inadequate.

RESPONSE:

See Static-O-Rfng Common Deficiency #7 and Response in Attachment B.


RESPONSE:


. . o

Page C-37

EQUIPMENT ITEM NUMBER: 70 EQUIPMENT DESCRIPTION: Static-O-Ring Pressure Switch Model 12NK4 LOCATION: Reactor Bldg., South Side, Elev. 962'6" PLANT ID NUMBERS: PS 5-12 (A-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:


Page C-38

EQUIPMENT ITEM NUMBER: 71 EQUIPMENT-DESCRIPTION: Barton Flow Switch Model 289 LOCATION: HPCI Room PLANT ID NUMBERS: FS 23-78 NRC QUALIFICATION CATEGORY: II.a



RESPONSE:


Page C-39

EQUIPMENT ITEM NUMBER: 72 EQUIPMENT DESCRIPTION: McDonnel Flow Switch Model AFIS LOCATION: SGTS Room PLANT ID NUMBERS: FS-2950, -2951 NRC QUALIFICATION CATEGORY: II.a



Specifically, no information was provided in the test report (FRC Reference 4208) to indicate that the test switches were representative of the installed switches, nor did the Licensee provide information

regarding similarity of the test switches to the installed switches.

RESPONSE:

FRC states, "The Licensee has not provided EDS file No. 26 rev. 3 for review." However, it should be noted that the FRC request for additional information (Appendix E to the TER) did not include this file. EDS calculation file 0910-001-26, Rev. 3, has been superseded by EDS calculation number 0910-106-EQ-16, Rev. D, dated 11 Nov 82. This new file was not available to the FRC at the time of the audit and addressed the deficiencies listed by the FRC as shown below.

Verified conversations with the manufacturer provides documentation that the limit switch installed in these flow switches are Micro Switch BA-1RV. Verified conversations with Micro Switch indicated that the only difference between a BA-lRV and a BA-2R Mi cro Switch is the switching mechanism and the contact gap. Therefore, the testing of Micro Switch Engineering Report 15027-1 (TER Reference 4208) is judged to apply to the BA-TRV.

Page C-40

EQUIPMENT ITEM NUMBER: 75 EQUIPMENT DESCRIPTION: Robertshaw Model SL4T2Al Level Switches LOCATION: Reactor Bldg., Elev. 935' W PLANT ID NUMBERS: LS 23-74, LS 23-75 NRC QUALIFICATION CATEGORY: II.a


T. "Adequate similarity between equipment and test specimen has not been established."

RESPONSE:

EDS File 0910-00T-EER-27 (TER Ref 40) was not intended to provide detailed qualification information; rather, it summarized qualification documented elsewhere. See also Generic Response item 6 in Attachment A.

Similarity between installed and tested equipment is now provided in EDS Calculation File 0910-106-EQ-22.

Correspondence from the manufacturer indicates that the following two differences exist between the test specimen and the installed equipment:

a. The tested unit used one ACRO microswitch model 286-0011-00, whereas the installed equipment uses two ACRO microswitches model 286-0028-00. The manufacturer considers this difference insignificant for the purpose of environmental qualification.

b. The tested unit used a V-groove gasket between the cover and base, whereas the installed equipment uses an 0-ring. This difference is also considered insignificant, as the sealing function is the same and because the installed equipment is not subject to the effects of chemical spray or flooding (submergence) where a totally sealed unit would be required.

Page C-41

EQUIPMENT ITEM NUMBER: 76 EQUIPMENT DESCRIPTION: Barton D/P Indicating Switch Model 288 LOCATION: Reactor Bldg., East Side, Elev. 935'O" PLANT ID NUMBERS: DPIS 13-83, DPIS 13-84, DPIS 116(A-D) thru 119(A-D) NRC QUALIFICATION CATEGORY: II.a

The correct model numbers for these items are as follows:

a. OPIS 13-83, DPIS T3-84-Barton Model 288.

b. DPIS 2-T16 (A-D) thru DPIS 2-119 (A-D)-Barton Model 278.

Barton Model 278 pressure switches differ from Model 288 pressure switches only by the use of mercury switches instead of snap-acting switches to perform trip functions. Because the mercury switches are built of mercury and metal electrodes encapsulated by glass, they are not susceptible to degradation from either radiation or thermal aging. The only accident parameter which could potentially affect the operation of these switches is pressure. However, the maximum pressure transient postulated for locations where Model 278 pressure switches are installed is 15.2 psia. This pressure transient falls within the band of atmospheric fluctuations. Although the change in pressure with respect to time is greater than for atmospheric fluctuations, the postulated transient is sufficiently low so as not to pose a credible failure for these switches.



RESPONSE:


Page C-42

EQUIPMENT ITEM NUMBER: 78 EQUIPMENT DESCRIPTION: Barton D/P Indicating Switch Model 288A LOCATION: Reactor Bldg., West Side, Elev. 935'0" PLANT ID NUMBERS: DPIS 23-76(A,B), DPIS 23-77(A,B) NRC QUALIFICATION CATEGORY: II.a


"Aging Degradation "Qualified Life or

Inadequately Evaluated." Replacement Schedule Has Not Been Established."

RESPONSE:

See Barton Common Deficiencies #2 and 3 in Attachment B.


RESPONSE:

See Barton Common Deficiency #4 in Attachment B.

4. "Required Profile Inadequately Enveloped."

RESPONSE:

See Barton Common Deficiency #5 in Attachment B.

5. Required Steam Exposure Inadequate.

RESPONSE:

See Barton Common Deficiencies #1 and 6 in Attachment B.

2. 2.

Page C-43

EQUIPMENT ITEM NUMBER: 81 EQUIPMENT DESCRIPTION: Yarway Level Indicator Model 4418C LOCATION: Reactor Bldg. PLANT ID NUMBERS: LIS 2-3-57(A,B), LIS 2-3-58(A,B), LIS 2-3-72(A-D) NRC QUALIFICATION CATEGORY: II.a


1. Inadequate similarity established between the tested and installed equipment.

RESPONSE:

See Yarway Common Deficiency #1 and Response in Attachment B.

2. Aging degradation inadequately evaluated. 3. Qualified life or replacement schedule not established. 4. Program to identify aging degradation not established. 5. Aging simulation not adequately regarded.

RESPONSE TO ITEMS 2, 3, 4 AND 5:

See Yarway Common Deficiencies #5, 6, 7, and 8, and Response in Attachment B.

6. Criteria regarding peak pressure inadequate.

RESPONSE:

See Yarway Common Deficiency #2 and Response in Attachment B.

Page C-44

EQUIPMENT ITEM NUMBER: 85 EQUIPMENT DESCRIPTION: Fenwal Temperature Switch Model 1700240 LOCATION: Steam Chase PLANT ID NUMBERS: TS2-121 122, 123, 124 (A-0) NRC QUALIFICATION CATEGORY: II.a


1. "Similarity Between Installed Equipment and Test Specimen Not Established."

RESPONSE:

See Fenwal Common Deficiency #1 and Response in Attachment B.


RESPONSE:

See FenwaT Common Deficiency #2 and Response in Attachment B.

3. Criteria Regarding Radiation Inadequate.

RESPONSE:


4. 5.

Aging Degradation Evaluated Inadequately. Qualified Life or Replacement Schedule Not Established.


See Fenwal Common Deficiencies #4 and 5 and Response in Attachment B.

Page C-45

EQUIPMENT ITEM NUMBER: 86 EQUIPMENT DESCRIPTION: Fenwal Temperature Switch Model 170236 LOCATION: HPCI Room PLANT ID NUMBERS: TS23-101, 102, 103, T04 (C-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:



RESPONSE:


4. Aging Degradation Evaluated Inadequately. 5. Qualified Life or Replacement Schedule Not Established.



Page C-46

EQUIPMENT ITEM NUMBER: 87 EQUIPMENT DESCRIPTION: FenwaT Temperature Switch Model 170236 LOCATION: Torus Compartment PLANT ID NUMBERS: TSl3-79, 80, 81, 82(A,B); TS23-101, 102, 103, 104(A,B) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:



RESPONSE:





Page C-47

EQUIPMENT ITEM NUMBER: 88 EQUIPMENT DESCRIPTION: Fenwal Temperature Switch LOCATION: RCIC Room PLANT ID NUMBERS: TS13-79, 80, 81, 82 (C-D) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:



RESPONSE:





Page C-48

EQUIPMENT ITEM NUMBER: 93 EQUIPMENT DESCRIPTION: Rosemount LOCATION: Torus Compartment PLANT ID NUMBERS: LT-7338 (A,B) NRC QUALIFICATION CATEGORY: II.a

Level Transmitter Model 1153DA5



RESPONSE:

See Rosemount Common Deficiency #1 in Attachment B.

2. 3.

Aging Degradation Evaluated Inadequately. Qualified Life or Replacement Schedule Not Established.

RESPONSE TO ITEMS 2 and 3:

See Rosemount Common Deficiencies #2 and 3 in Attachment B.

4. Criteria Regarding Steam Exposure Inadequate.

RESPONSE:


5. 6. 7.

Criteria Regarding Radiation Not Satisfied. Criteria Regarding Test Failure or Severe Anomalies Not Satisfied. Criteria Regarding Instrument Accuracy Not Satisfied.

RESPONSE TO ITEMS 5, 6 AND 7:

See Rosemount Common Deficiencies #5, 6, and 7 in Attachment B.

Page C-49

EQUIPMENT ITEM NUMBER: 99 EQUIPMENT DESCRIPTION: GE Electric Motor Model 5K254AK205 LOCATION: SGTS Room PLANT ID NUMBERS: V-EF-17 (A,B) NRC QUALIFICATION CATEGORY: II.a


1. Documented Evidence of Qualification Inadequate.

RESPONSE:

See Generic Response item 2 in Attachment A.

EDS calculation No. 0910-106-GEMOT-12 re-evaluates GE motors. It is considered that this new file will meet the FRC concerns of providing a comprehensive analysis of the GE motors that could be verified by an independent reviewer.

2. 3.

Aging Degradation Inadequately Evaluated. Qualified Life or Replacement Scheule Not Established.


See GE Motors Common Deficiencies #1 and 2 and Response in Attachment B.

4. Radiation Criteria Not Satisfied.

RESPONSE:

EDS calculation No. 0910-106-GEMOT-l2 re-evaluated GE motor analysis of vendor-supplied materials lists and reduced the tism of previous evaluations. This resulted in a radiation of 1x10 6 Rads gamma, which is greater than the 8.7x10 5 Rads radiation dose specified for this equipment.

radiation conservathreshold gamma

Page C-50

EQUIPMENT ITEM NUMBER: EQUIPMENT DESCRIPTION: LOCATION: RHR Room PLANT ID NUMBERS: P-2 NRC QUALIFICATION CATE

100 GE Electric Motor Model 5K6329XC4A

02 (A-D) GORY: II.a


1. Similarity between test specimen established.

and equipment inadequately

RESPONSE:

EDS calculation No. 0910-106-GEMOT-12 re-evaluated GE motors with a report supplied to Northern States Power by GE specifically for the purpose of environmental qualification of the installed motors.

Aging degradation evaluated inadeuateTy. Qualified life not established.



2. 3.

EQUIPMENT ITEM NUMBER: EQUIPMENT DESCRIPTION: LOCATION: RHR Room PLANT ID NUMBERS: P-2 NRC QUALIFICATION CATE

101 GE Electric Motor Model 5K6338XC298

08 (A,B) GORY: II.a


1. Similarity between test specimen and equipment inadequately established.

RESPONSE:

EDS calculation No. 0970-T06-GEMOT-12 re-evaluated GE motors with a report supplied to Northern States Power by GE specifically for the purpose of environmental qualification of the installed motors.

2. 3.

Aging degradation evaluated inadequately. Qualified life not established.



Page C-5T

Page C-52

EQUIPMENT ITEM NUMBER: 113 EQUIPMENT DESCRIPTION: Allen-Bradley Terminal Board Model 1492-CD3 LOCATION: SGTS Room PLANT ID NUMBERS: None NRC QUALIFICATION CATEGORY: II.a


1. "Radiation criteria not satisfied."

Specifically, there that data on DuPont board made of Fosta

is insufficient information presented to establish ZyteT films would be applicable to a terminal 512.

RESPONSE:

Allen-Bradley Model 1492-CD3 Terminal Boards have been re-evaluated in EDS calculation file No. 0910-T06-EQ-01. The re-evaluation is based on the results of LOCA testing documented in EDS Report 04-0910-54, Revision 0, March 1983.

Stage 2 of this testing (Environmental Qualification Testing) included the following segments:

a. b. c. d.

Radiation aging/Functional testing - 4x10 5 Rads Temperature aging/Functional testing Thermal and humidity cycling/Functional testing Accident radiation testing/Functional testing - 2.5x108 Rads.

Subsequent LOCA testing was not considered for the evaluation of this equipment since radiation is the only harsh accident parameter.

Since the maximum postulated accident dose is 3x10 6 Rads, the re-evaluation of this equipment based on the testing above qualifies them for operation at installed locations.

Page C-53

EQUIPMENT ITEM NUMBER: T20 EQUIPMENT DESCRIPTION: Fisher E/P Transducer Model 546 LOCATION: RHR Pump Room PLANT ID NUMBERS: E/P 1728, E/P 1729 NRC QUALIFICATION CATEGORY: II.a


1. "Documented Evidence of Qualification Inadequate." Specifically, manufacturer has disclaimed the suitability of the equipment for operability after an accident, and the documentation provided in EDS Report 17 does not address the vendor statement.

Reference cited: Letter from J.S. McLagan (Fisher Controls Company) to S.R. Christensen (Portland General Electric), dated 3-2-81 (TER PGR 4718).

RESPONSE:

The following paragraph was quoted from the above Fisher Controls Center.

"It is important to emphasize that this summary is not intended to demonstrate suitability of the 546 for operability after an accident, but only to show performance characteristics exhibited in previous tests. As mentioned in this document, no further testing or pursuit of TE qualification is contemplated at this time."

From the above paragraph it is clear that Fisher controls has not "disclaimed the suitability of the equipment for operability after an accident" as claimed by the FRC. The Fisher Controls Tetter states that tests performed on their equipment, and summarized in their report, should not be taken as a per se qualification of the Fisher type 546 transducer as Class TE equipment; however, it does not say that the test results cannot be used with supplemental qualification data to establish qualification for this equipment in a comprehensive qualification evaluation.

Page C-54

EQUIPMENT ITEM NUMBER: 121 EQUIPMENT DESCRIPTION: Hea LOCATION: SGTS Room PLANT ID NUMBERS: Ti NRC QUALIFICATION CATEGORY:

The FRC reviewed calculation file dated 13 Dec 82, new file was not the deficiencies

vy Duty Transformer Model SZO

I.b

EDS calculation No. 0910-001-21, Rev. 3, 11 Nov 81. This has been superseded by EDS calculation No. 0910-106-EQ-12, in which the Heavy Duty transformer was re-evaluated. This available to the FRC at the time of the audit and addresses listed by the FRC as shown below.


1. "Inadequate Aging Degradation Evaluation."

RESPONSE:

See Generic Response for item 1 in Attachment A.

EDS calculation No. 0910-106-EQ-13 re-evaluated the transformer and has entered the qualified thermal life to greater than 40 years, as determined by analysis of a vendor-supplied materials list.

2. "Replacement Schedule Not Established."

RESPONSE:

Replacement Schedule not required, as qualified thermal life is greater than 40 years.

3. "Radiation Criteria Not Satisfied".

RESPONSE:

EDS calculation No. 0910-T06-EQ-73 re-evaluated radiation and concluded the minimum radiation threshold is 8.2 x 105 Rads gamma, which fully envelopes the specified 5 x 105 Rads gamma dose.

Page C-55

EQUIPMENT ITEM NUMBER: 125 EQUIPMENT DESCRIPTION: Woodward Governor EG-R Model R8250133 LOCATION: HPCI Room PLANT ID NUMBERS: HPCI Turbine: Speed Control NRC QUALIFICATION CATEGORY: II.a

The FRC reviewed EDS calculation No. 0910-001-SER-35. The Woodward Governor turbine controls have been re-evaluated in EDS calculation No. 0910-T06-EQ-27, dated 3 Feb 83. This new file was not available to the FRC at the time of the audit and addresses .the deficiencies listed by the FRC as shown below.

It is considered that the new file will meet the FRC concerns of providing a comprehensive analysis that can be verified by an independent reviewer.



RESPONSE:

The original submittal for this equipment item was not complete, pending receipt of more detailed information from the manufacturer. This information has been received and evaluated fn EDS calculation No. 0910-106-EQ-27. Qualification has been established in this calculation file for all parameters except radiation, in that the solenoid leadwire insulation contains teflon. Northern States Power Company will replace this solenoid with a radiation-resistant model.

Page C-56

EQUIPMENT ITEM NUMBER: 129 EQUIPMENT DESCRIPTION: GE Electrical Penetration Model NSO-4 LOCATION: Containment PLANT ID NUMBERS: JX-105 (A,C,D) NRC QUALIFICATION CATEGORY: II.a

EDS calculation No. 0910-106-GEPEN-12 re-evaluates the equipment. This new file was not available to the FRC at the time of the audit and addresses similarity as shown.


1. "Similarity Between Equipment and Test Specimen Not Adequately Established."

RESPONSE:

The following is taken from EDS calculation No. 0910-106-GEPEN-12:

"6.1 Test Specimen Evaluation

"The NS02, NS03, and NS04 containment penetrations are all generically equivalent canister-type penetrations according to Reference 25. The FOT Electrical penetration described in Reference 5 corresponds with and is identical to the NS02-type canister (25). The only difference in the NSO2, NSO3, and NS04 penetrations is in the type of electrical cable used in each assembly (25). The NS04 penetration, to be evaluated here, is low voltage for 600V power control and instrumentation leads (25); although the General Electric "FOT Electrical Penetration Assembly" test report specifically addresses the NSO2 penetration, it is also applicable to the NSO4 penetration (excluding the cable), since they are both equivalent canister-type penetrations.

"General Electric's "Electrical Penetration Assemblies Prototype Testing Qualification Report" (6), is directly applicable to canister-type penetrations as stated in D.L. Butcher's (GE) letter to N. Smith (CECo) dated 3-6-79 (6). Since the NS04 is a canister-type penetration, the testing done in Reference 6 is applicable."

Page C-57

REFERENCES:

(5) "Qualification Test For FOT Electrical Penetration Assembly," by R.M. Schuster, dated 4-30-71, EDS DI 0910-001-030

(6) Letter to N. Smith (CECo) from D.C. Butcher (GE) dated 3-6-79, with attached test report, "Electrical Penetration Assemblies Prototype Testing Qualification Report," by J.P. Watson, dated 3-16-70, EDS DI 0910-001-097

(25) Letter to L.C. Lessor (NPPD) from R.H. Lube (GE) 72/7/77, Re: Containment penetration test report, EDS DI #0970-106-086.

Page C-58

EQUIPMENT ITEM NUMBER: 131 EQUIPMENT DESCRIPTION: GE Electrical Cable Model SI58042 LOCATION: Various PLANT ID NUMBERS: None NRC QUALIFICATION CATEGORY: II.a



Page C-59

EQUIPMENT ITEM NUMBER: 132 EQUIPMENT DESCRIPTION: GE Electrical Cable Model SI58170 LOCATION: SGTS Room PLANT ID NUMBERS: None NRC QUALIFICATION CATEGORY: II.a



b

Page C-60

EQUIPMENT ITEM NUMBER: 133 EQUIPMENT DESCRIPTION: GE Electrical Cable Model SI58175 LOCATION: SGTS Room PLANT ID NUMBERS: None NRC QUALIFICATION CATEGORY: II.a



Page C-61

EQUIPMENT ITEM NUMBER: 135 EQUIPMENT DESCRIPTION: GE Electrical Cable Model SI58007 LOCATICN: Steam Chase PLANT ID NUMBERS: None NRC QUALIFICATION CATEGORY: II.a


1. "Similarity Between Equipment and Test Specimen Inadequately Established."

RESPONSE:


These cables have been re-evaluated in EDS calculation No. 0910-106-GECAB-1 2.

Wyle test report #44114-1 tested General Electric SI-58007 cable taken from the Monticello plant. This testing provides adequate documentation to qualify this calbe for all parameters except thermal aging.

Another Wyle test report, #44114-2, tested GE SI-58136 cables, which are made of the same materials as the SI-58007. This test report provides adequate documentation of the ability of these materials to withstand degradation from thermal aging.

Page C-62

EQUIPMENT ITEM NUMBER: 136 EQUIPMENT DESCRIPTION: GE Electrical Cable Model S158007 LOCATION: Outside Containment PLANT ID NUMBERS: None NRC QUALIFICATION CATEGORY: II.a


1. "Similarity Between Equipment and Test Specimen Inadequately Establ i shed."

RESPONSE:


These cables have been re-evaluated in EDS calculation No. 0910-106-GECAB-1 2.

Wyle test report #44114-1 tested General Electric SI-58007 cable taken from the Monticello plant. This testing provides adequate documentation to qualify this calbe for all parameters except thermal aging.

Another Wyle test report, #44114-2, tested GE SI-58136 cables, which are made of the same materials as the SI-58007. This test report provides adequate documentation of the ability of these materials to withstand degradation from thermal aging.

Page C-63

EQUIPMENT ITEM NUMBER: 140 EQUIPMENT DESCRIPTION: GE Electrical Cable Model S158136 LOCATION: Various PLANT ID NUMBERS: None NRC QUALIFICATION CATEGORY: II.a


1. Criteria Regarding Radiation Not Satisfied. 2. Criteria Regarding Test Failures Not Satisfied.

RESPONSE:

These deficiencies result from using Wyle Test Report #44114-2 to document qualification. These cables have been re-evaluated in EDS calculation No. 0910-106-GECAB-12. This re-evaluation cites another Wyle test report, #44114-1 , as documentation for all parameters except radiation, thermal aging, and operating time.

Wyle test report #44114-2 exposed GE cable SI-58136 to 4.0 x 107 Rads gamma. The Wyle Report states that the cable insulation became tarlike after this irradiation. The extreme damage to the cable insulation by the irradiation caused the subsequent failure of this cable. Since the Maximum Accident Dose to GE cable SI-59136 is only 3.0 x 106 Rads gamma, the Wyle irradiation test of the cable to 4.0 x 107 Rads is excessive and does not conclusively prove the cable will not function after the accident dose. Therefore, a materials analysis was performed to determine the maximum radiation dose that GE cable SI-58136 can withstand. A butyl-rubber-insulated neoprenejacketed cable (identical to the materials of construction in GE cable SI-58136) was tested to determine maximum serviceable radiation dose as reported in "Insulation and Jackets for Control and Power Cables in Thermal Reactor Nuclear Generating Stations", R. B. Blodgett and R. G. Fisher, EDS DI #0910-001-106. From this report the maximum serviceable radiation dose was determined to be 5.0 x 106 Rads gamma. EPRI Report NP-2129, "Radiation Effects on Organic Materials in Nuclear Plants", cites several other references which indicate that the limits for butyl-based insulation is 5 x 106 Rads.

Since the maximum dose specified for this cable is 3 x 106 Rads, and because the radiation resistance of neoprene is better than that for butyl rubber, these cables are considered acceptable for operation in specified locations.

Page C-64

EQUIPMENT ITEM NUMBER: 142 EQUIPMENT DESCRIPTION: Carol Wire and Cable Model 1050 LOCATION: SGTS Room PLANT ID NUMBERS: None NRC QUALIFICATION CATEGORY: II.a


T. "Criteria Regarding Radiation Is Not Satisfied. Specifically, 'The reference to the Cannon Electrical Report for PVC is not adequate for qualification of cable'."

RESPONSE:

Several references are available that PVC insulated cables can withstand the postulated accident radiation dose of 3.0 x 106 Rads gamma. The most important properties that the cable must retain are dielectric and mechanical strength. From the paper "Insulations and Jackets for Control and Power Cables in Thermal Reactor Nuclear Generating Stations", by R.B. Blodgett and R.G. Fisher, IEEE Transactions on Power Apparatus and Systems, May 1969, PVC insulation increases in tensile strength up to 5.0 x 106 Rads gamma. Percent elongation also increases slightly up to 5.0 x 106 gamma. Retention of insulation resistance is also proven, since the dielectric strength constant increases up to 1.0 x TO Rads gamma.

It is stated in "Radiation Resistance of Cable Insulating Materials for Nuclear Power Generating Stations", by Kuriyama, Hayakawa, Nakase, published in IEEE Transactions Electric Insulation, June 1978, that PVC insulation is utilizable even at 108 Rads gamma irradiation single-environment.

From Sandia National Laboratories Report SAND80-2149, "RadiationThermal Degradation of PE and PVC: Mechanism of Synergism and Dose Rate Effects", by R.L. Clough and K.T. Gillen, it was determined that 80% of the initial elongation was retained after 7.9 x 106 Rads gamma irradiation.

From EPRI's report NP-2129, "Radiation Effects on Organic Materials in Nuclear Plants", Figure 3-1, data is presented on the effect of irradiation on PVC cables from 38 manufacturers. This data shows that PVC cable will retain 75-80 per cent of its initial elongation after 5.0 x 107 Rads gamma.

h

Page C-65

Since there is no mechanical Toad on the.cable, the small reduction in per cent elongation after irradiation at 7.9 x 106 Rads gamma or higher is inconsequential. In addition, it has been shown that electrical insulation strength is maintained up to 1.0 x 108 Rads gamma. The only harsh environmental parameter that Carol wire and cable model 1050 will be subject to is radiation. Since all of the above data envelops the required accident radiation dose of 3.0 x 106 Rads, the cable is qualified.

Page C-66

EQUIPMENT ITEM NUMBER: 144 EQUIPMENT DESCRIPTION: Samuel Moore Electrical Cable Model 1952 LOCATION: Reactor Bldg., Elev. 962'O" PLANT ID NUMBERS: Electrical Distribution NRC QUALIFICATION CATEGORY: II.a



RESPONSE:

Similarity for this item was established by comparing manufacturers specifications for installed cables with those for samples #7, 13, and 14 in the cited test report. The installed cables are identical to those tested on the basis of design and materials of construction.

2. "Aging Degradation Not Evaluated Adequately." 3. "Qualified Life or Replacement Schedule Not Established."


See Generic Response for item I in Attachment A.

Thermal aging for this equipment has been addressed in EDS Calculation 0910-106-EQ-29. The qualified life for these cables is greater than 40 years based on pre-accident aging tests documented in "Qualification Testing of Electrical Cables under simulated LOCA/DBE Conditions." This testing was performed for Samuel Moore Dekoran Division by Isomedix Inc., June 1978.

Page C-67

EQUIPMENT ITEM NUMBER: 145 EQUIPMENT DESCRIPTION: General Atomic Radiation Detector Model RD23 LOCATION: Drywell, Elev. 948' PLANT ID NUMBERS: RE-7860 (A,B) NRC QUALIFICATION CATEGORY: II.a


1. Adequate Similarity Between Equipment and Test Specimen not Establ i shed.

RESPONSE:

The installation at Monticello uses technique #4 (Raychem heat shrink tubing without silicone grease loading) as described in General Atomic test report E-254-960 (Revised 5/1/81). This configuration successfully passed the LOCA environment testing.



As described in General Atomic Test Report R-254-960 (revised 5/1/81), the components of the RD-23 detector assembly were evaluated and determined to not be susceptible to thermal aging (e.g., Reuter Stokes detector) or were previously qualified by the vendor for a 40-year life (e.g., Rockbestos coaxial cable, Raychem WCSF-N heat shrink tubing). Thus, thermal aging degradation has been adequately considered.

4. Criteria Regarding Radiation Not Satisfied.

RESPCNSE:

As described in General Atomic Test Report R-254-960 (revised 5/1/81), the components of the RD-23 detector assembly were evaluated and determined to not be susceptible to radiation effects or were previously qualified by the vendor to a dosage of 2x10 8 Rads. Thus, the radiation qualification requirement is satisfied.

-J

Page C-68

EQUIPMENT -ITEM NUMBER: 146 EQUIPMENT DESCRIPTION: Pressure Controls Pressure Switch Model A171P LOCATION: Drywell PLANT ID NUMBERS: PS-7356, 7358, 7359, 7307, 7463, 7464, 7465, 7466 NRC QUALIFICATION CATEGORY: II.a




The basis and justification for the pre-aging performed during the qualification test is contained in Wyle Laboratories Qualification Plan 57571, Rev. B, which is available for review upon request.

3. Criteria Regarding Spray Not Satisfied.

RESPONSE:

The switch housing is a NEMA 4 design which is designed to withstand direct spray impingement.

Furthermore, as part of another plant modification, these pressure switches will be removed and replaced with qualified transmitters located outside containment.

Page C-69

EQUIPMENT ITEM NUMBER: 147 EQUIPMENT DESCRIPTION: Rosemount Pressure Transmitter Model 1153AA17 LOCATION: Reactor Bldg, ETev. 985', 962' PLANT ID NUMBERS: PT-7251 (A,B) NRC QUALIFICATION CATEGORY: II.a



RESPONSE:

See Generic Response for item 2 in Attachment A.

2. Aging degradation evaluated inadequately. 3. Qualified life or replacement schedule not established.


See Rosemount Common Deficiencies #2 and 3 in Attachment B.

4. Criteria regarding steam exposure inadequate.

RESPONSE:


5. Criteria regarding radiation not satisfied. 6. Criteria regarding test failure or severe anomalies not satisfied. 7. Criteria regarding instrument accuracy not satisfied.


See Rosemount Common Deficiencies #5, 6, and 7 in Attachment B.

Page C-70


with Class B, DC motors; Motor Brake installed LOCATION: HPCI Room PLANT ID NUMBERS: MO-2067, 2071 NRC QUALIFICATION CATEGORY: II.a



RESPONSE:



RESPONSE:



RESPONSE TO ITEMS 3 AND 4: 1

See Limitorque Common Deficiencies #3 and 4 and Response in Attachment B.

Page C-71


with Class B, DC motors LOCATION: Steam Chase PLANT ID NUMBERS: MO-2107 NRC QUALIFICATION CATEGORY: II.a



RESPONSE:


2. "Adequate Similarity Between Equipment and Test Specimen Not Establ i shed."

RESPONSE:




See Limitorque Common Deficiencies #3 and #4 and Response in Attachment B.

Page C-72


with Class B, DC motors LOCATION: Reactor Bldg., Elev. 948' PLANT ID NUMBERS: MO-2014, 2015 NRC QUALIFICATION CATEGORY: II.c




See Limitorque Common Deficiencies #3 and #4 and Response in Attachment B.

Documents

Forwards addl info & justification to support resolution ... · qualification using methods other than strict application of data obtained from type testing in accordance with either