ML18078A962
| ML18078A962 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 03/06/1979 |
| From: | Mittl R Public Service Enterprise Group |
| To: | Parr O Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7903120233 | |
| Download: ML18078A962 (34) | |
Text
PS~G Public Service Electric and Gas Company 80 Park Place Newark, N.J. 07101 Phone 201 /430-7000 March 6, 1979 Director of 2Nuclear Reactor Regulation
- u. s. Nuclear Regulatory Comm1ssion Washington, D. c.
20555 Attention:
Gentlemen:
Mr. Olan D. Parr, Chief Light Water Reactors Branch 3 Division of Project Management RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 Public Service Electric and Gas Company hereby submits sixty (60) copies of its response to your request for additional in-formation regarding sufficient auxiliary feedwater in the event of a tornado missile strike Question 1.41, Equipment Qualification Question 41.1 parts 1 thru 8 and updated pages to FSA~ Appendix D.
Should you have any questions, please do not hesitate to contact us.
Attachment 790312023'5 The Energy People t!d~rs, R. L. Mittl General Manager -
Licensing and Environment Engineering and Construction 95-2001 (400M) 9-77
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Public Service Electric and Gas Company 80 Park Place Newark. N.J. 07101 Phone 201 '430-7000 March 6, 1979 Director of Nuclear Reactor Regulation
- u. s. Nuclear Regulatory Commission Washington, D. c.
20555 Attention:
Gentlemen:
Mr. Olan D. Parr, Chief Light Water Reactors Branch 3 Division of Project Management RESPONSE TO REQUESTS FOR ADDITIONAL INFORMATION NO. 2 UNIT SALEM NUCLEAR GENERATING STATION DOCKET NO. 50-311 Public Service Electric and Gas Company hereby submits sixty (60) copies of its response to your request for additional in-formation regarding sufficient auxiliary feedwater in the event of a tornado missile strike Question 1.41, Equipment Qualification Question 41.1 parts 1 thru 8 and updated pages to FSAR Appendix D.
~
Should you have any questions, please do not hesitate to contact us.
Attachment The Energy People
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Ver~y t1u{~iftf~*
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R. L. Mittl General Manager -
Licensing and Environment Engineering and Construction
'""°' ""°'" ':"_____
QUESTION 1.41 Supplement the discussion in the FSAR response to Question 5.16 concerning tornado protection.
Describe the capability and procedures for bringing the plant to a safe shutdown condition in the event of a design basis tornado.
Water tanks and systems not specifically designed to withstand tornado. induced missiles cannot be assumed to be available for use.
Include in your description such things as auto-matic water supply switchover and technical specifications on manual switchover hardware.
ANSWER The response to Question 5.16 describes in detail the various primary and backup water sources available to the Auxiliary Feedwater System and the Chemical and Volume Control System for use in attaining and maintaining safe shutdown.
The backup water sources from the eves holdup tanks, the spent fuel pool and the Service Water System are
- located in buildings/structures designed to withstand tor-nado induced missiles.
The water storage tanks described in the response to Question 5.16 are located in different areas of the station site (refer to FSAR Figure 1.2-1).
Although not specifically designed to withstand tornado induced missiles, the separated locations of the various tanks precludes the possibility that all tanks would be rendered unavailable due to tornado induced missiles.
The systems which are required to bring the unit to a safe shutdown and enclosed in tornado protected buildings/
structures and capable of being powered by the Standby AC Ql. 41-1 M P79' 03 11
Power System.
In the unlikely event that a breach of any of the primary water sources were to occur (Refueling Water Storage Tank, Auxiliary Feedwater Storage Tank or Primary Water Storage Tank), station operating personnel would initiate procedures to bring the unit to a hot shutdown condition.
Hot shutdown can be maintained with Auxiliary Feedwater System operation and main steam atmospheric relief valves.
Boration can be accomplished via the charging pumps, boric acid tanks and boric acid transfer pumps.
The backup water source for the Auxiliary Feedwater System is dependent on which storage tank *is rendered unavailable.
Transfer to a backup water source is accomplished with remote manual valves (operable from the Control Room) except for the service water backup source, which requires installation of a spool piece as described in the response to Question 5.16.
No automatic water supply switchover is provided.
The backup water sources for the Auxiliary Feedwater System water sources are addressed in the Technical Specifications for No. 1 Unit (except for the service water backup source)
- The unit would be maintained in a hot shutdown condition until station operating personnel have assessed operating Ql.41-2 M P79 03 12
It has been determined by actual demonstration that two men can install the spool piece in 13 minutes.
The installation requires no special tools and involves removing two blind flanges and bolting the spool piece in place.
The spool piece is stored at the connection point.
An analysis was performed to determine the time period following a loss of AC power and main and auxiliary feedwater flow before the core becomes uncovered.
The pertinent assumptions used in the analysis are as follows:
All AC power lost at time of incident Rods assumed to begin dropping into core 2 seconds following incident ANS standard decay heat curve assumed Pressurizer relief and safety valves operative Initial power is 1.02 times rated power Loss of all main and auxiliary feedwater following incident.
The calculation was subdivided into the following areas:
- a. Heat required to raise primary to saturation temperature
- b. Heat required to uncover core after primary saturation temperature is reached.
Ql.41-4 M P79 03 14
- c. Secondary heat sink available Primary Saturation The amount of heat required to raise the primary inventory from the initial average temperature to the saturation temperature corresponding to the pressurizer power operated relief valve setpoint was calculated to be approximately 16 full power seconds.
Core Uncovering The quantity of heat required to boil sufficient primary inventory in order to begin uncovering the core was found to be approximately 5.8 full power seconds.
Secondary Heat Sink Based upon nominal initial secondary mass, the heat required to deplete the secondary inventory via the steam generator safety valves was calculated to be equivalent to 74.8 full power seconds.
The total heat generation which results in uncovering the core is then the sum of the three heat sinks itemized above.
These components yield a total of 96.6 full power seconds.
Assuming a standard ANS decay heat curve, the period of time following initiation of the incident was calculated to be approximately 4200 seconds or 70 minutes.
Ql.41-5 M P79 03 15
Assuming a standard ANS decay heat curve, the period of time following initiation was calculated to be approxi-mately 4200 seconds or 70 minutes.
With the above analysis analysis taken into consideration, a conservative estimate of time history to initiate auxiliary feedwater following a postulated loss of outdoor water storage tanks is provided below:
Time (Min.)
0 30 40 53 Event Tornado induces failure in all out-door water storage tanks.
(Low water indication in Control Room)
Acknowledgement by Control Room, dis-patch personnel to install spool piece.
Spool piece installation commences Spool piece installation complete, auxiliary feedwater initiated.
The above discussion demonstrates that sufficient water sources are available to bring the unit to a safe shutdown condition in the event of a breach of a primary water source due to a tornado induced missile.
Ql.41-6 M P79 03 16
QUESTION 7.41 (1)
The MSLB accident environmental envelop inside the contain-ment should be provided to verify the adequacy of the quali-fication of those Class IE equipment required to mitigate this event.
ANSWER The MSLB analysis for Salem was presented in the response to QS.82.
The analysis indicates that the worst case peak temperature is 333.SoF.
Table Q7.30 has been revised to include the MSLB environmental conditions.
All equipment required to functionally operate during an MSLB event will be demonstrated to be qualified to a peak temperature of 3SOOF.
In accordance with information received from the NRC in late December 1978 concerning the Salem MSLB analysis, the peak temperature profile for this event will encompass a temperature above 300°F for three minutes with at least one minute at 3500F.
SNGS-FSAR UNITS 1 & 2 MP79608 l Q7.41.l (1)
QUESTION 7.41(2)
In your response you stated that Barton Model 763 and 764 transmitters were found acceptable on the North Anna plant.
We orally informed you that the review of these transmitters is being done as part of our review of the Westinghouse reverification program, and that this review has not been completed, and that NRC gave only the North Anna plant for an interim period only.
Provide the information necessary to demonstrate that the Barton transmitters are acceptable for use for an interim period at Salem Unit 2.
ANSWER The response to Q7.30 has been revised to correct the statement that the Barton 763 and 764 model transmitters were found acceptable on the North Anna plant.
Salem will use these transmitters for the following functions:
Pressurizer Level (Barton 764)
Steam Generator Level (Barton 764)
Reactor Coolant System Wide Range Pressure (Barton 763)
The qualification data for the Barton transmitters is presented in the Westinghouse letter to the NRC of September 29, 1978 (NS-TMA-1950).
In this letter report both a summary of the equipment qualifications and test results are presented.
The information contained in this report is proprietary information of Westinghouse Electric Corporation.
This report addresses the concerns of the NRC rasied during the North Anna, D. c. Cook and Diablo Canyon Licensing Review.
The information contained in the report SNGS-FSAR UNITS 1 & 2 Q7.41(2)-l MP79609 1
is applicable for the Salem Plant and demonstrates the acceptable use of the Barton Transmitters.
The Barton Transmitters are enclosed in instrument panels which afford physical protection and a thermal barrier.
Additional confirmatory test data will be prepared on instrument panel assemblies and the enclosed Barton Transmitters to demonstrate acceptability for a Salem MSLB/
LOCA environment.
Refer to the response to Q.741(6) for test results.
SNGS-FSAR UNITS 1 & 2 MP79609 2 Q7.41(2)-2
QUESTION 7.41(3)
The qualification data for the Rosemount transmitter (Model 1153A) was provided in a Rosemount report (No. 3788) which was first submitted on the AN0-2 docket and referenced by you.
We have found the following two-problems in this test report.
(a)
The temperature profile for the saturated steam test shows a rise time of about 3 minutes from room temperature to the maximum temperature.
However, the typical rise time at Salem plant under LOCA is about 10 to 20 seconds.
Therefore, demonstrate how the test results in Rosemont report No. 3788 apply to Salem Unit
- 2.
(b)
The test setup and the actual field installation are not the same.
The installation interface wa~ not tested.
The AN0-2 applicant provided additional test data to resolve these concerns.
We require that you provide additional information (as was furnished on AN0-2 docket) to resolv~ the above stated concern.
ANSWER (a)
The peak temperature of the Rosemount Model 1153A transmitters test report is 3SOOF.
This peak temper-ature is compatible with MSLB analysis as described in the response to Q7.41(1).
A typical performance profile form Rosemount Report No. 3788 indicates that the transmitter.test temperature rose approximately 1200F to 3000F in less than two minutes with the total rise time to the peak temperature of 350° in approximately 2-1/2 minutes.
The 3S0°F temperature was held for over 10 minutes.
The satisfactory results of this test indicates the acceptability of the instrument for a Salem MSLB event.
SNGS-FSAR UNITS 1 & 2 MP79610 1 Q7.41(3)-l
The peak temperature for a LOCA at Salem has been determined to be 2710F.
This peak temperature occurs in approximately 10 to 20 seconds.
An examination of the Rosemount report indicates that this temperature was achieved during the test in a little over one minute.
The Arkansas Power and Light Company has provided further data on the capability of the Rosemount 1153A transmitters in their letters to the NRC dated 9/15/78, 9/26/78 and 10/16/78 for the Arkansas Nuclear One -
Unit 2 application.
Their test data demonstrates that the Rosemount transmitter will work properly with rise times on the order of 10-20 seconds.
Their accident transient indicates a rise time of approximately 12.5 seconds to reach a temperature of 2700F.
In their letters they have provided data to show the applicability of the test response with the accident transient.
Their satisfactory qualification data indicates the acceptability of the instrument for a Salem LOCA event.
The ANO seismic and integrated radiation doses are applicable for the instrument applications at Salem.
The peak temperature of the Arkansas data exceeded 3oooF.
The Salem MSLB analysis indicates a peak temperature of 350°F.
The Virginia Electic Power SNGS-FSAR UNITS 1 & 2 Q7.41(3)-2 MP79610 2
Company has performed thermal transient analyses on a number of components for the North Anna Station.
The results of these analyses were presented in response to Comment 7.17 on the North Anna appiication.
In this response they provided thermal analysis data for a Rosemount transmitter.
The analysis was done in accordance with NRC CSB Branch Technical Position 6~2.
A simple model was utilized for the transmitter only taking into account the electronics housing.
Although done for a different model Rosemount transmitter, the model utilized for the electronics housing is similar to the 1153A transmitter.
Their analysis was ba.sed on a peak temperature of 4400F, a more severe transient than 350°F.
The results of the anlaysis indicated a peak surface temperature of 264°F which is considerably less than the qualification temperature of Arkansas (greater than 3QQOF).
This provides further information to demonstrate the suitability of the 1153A transmitter for use at Salem.
The Rosemount transmitters are enclosed in Instrument Panels which afford Physical Protection and a Thermal Barrier.
Additional Confirmatory Test Data will be prepared on an Instrument Panel assemblies and the enclosed Rosemount Transmitters to demonstrate acceptability for a Salem MSLB/LOCA environment.
Refer to response to Q_7. 41 (6).
SNGS-FSAR UNITS 1 & 2 MP79610 3 Q7.41(3)-3
(b)
The electrical connection interface with the Rosemount 1153A transmitter will be through a Conax stainless steel packing gland.
The electrical wires will feed through a Grafoil (trademark of Union Carbide) grommet in the packing gland.
The radiation resistance of stainless steel is well known.
Grafoil has an excellent resistance to radiation (l.5xlo9 rads) and a temperature range of -400°F to +soooF which are well within the application criteria for Salem.
In addition, the transmitter and interface are enclosed in an instrument panel so that they are not physically exposed to the caustic sprayo This interface should assure operability of the transmitter for the environmental conditions at Salem.
Additional confirmatory test data will be prepared on the interface connection for use at Salem.
Refer to response to Q7.41(6).
SNGS-FSAR UNITS 1 & 2 MP79610 4 Q7.41(3)-4
The elctrical connection interface will be similar to that proposed for the Rosemount transmitters described in response to Q7.41(3) Part b.
Additional confirmatory test data on initial thermal transients will be prepared on the ASCO solenoid valves used at Salem.
Refer to response to Q7.41(6).
SNGS-FSAR UNITS 1 & 2 Q7.41(4)~2 MP79611 2
-~--
QUESTION Q7.41(5)
We require that you submit the NAMCO report covering limit switches that was referenced in the FSAR.
It our position that limit switches for containment isolation valve position indication should be qualified to the most severe containment environment.
We require that you identify and justify the use of those limit switches that will not meet this qualification requirement.
ANSWER Copies of the qualification report for NAMCO limit switch model EA-180 will be forwarded for your review.
These limit switches are being used for in-containment isolation valves as described in our response to Q7.35 which addresses NRC IE Bulletin 78-04.
The response to Q7.35 addresses isolation valve position indication and the use of qualified limit switches.
The electrical connection interface will be determined on the basis of the Instrument Panel test.
Refe~ to response toQ7.4l(n).
SNGS-FSAR UNITS 1 & 2 MP79612 1 Q7.41(5)
QUESTION Q7.41(6)
We require that you submit for our review an environmental qualification test plan and test results for instrument panels located inside the containment for transmitters and solenoids enclosure.
ANSWER The following information encompasses the test plan for an instrument panel qualification test:
Purpose Instrument transmitters, most solenoids and other miscellaneous items are enclosed in instrument panels at Salem.
Safety-related instruments and solenoids for in-containment applications have been environmentally qualified by various manufacturers, suppliers and/or users.
In some cases this equipment qualification does not totally encompass the conditions postulated at Salem for peak temperatures or fast temperature rise times.
The purpose of the instrument panel test is to provide data to show that the panel provides thermal protection for the enclosed instruments.
Objective The objective of this test is not to repeat type tests that have already been performed, but to augment them with data obtained from a Salem environmental test profile.
SNGS-FSAR UNITS 1 & 2 MP79613 1 Q7.41(6)-l
- 1)
Determine if the protection provided to components by the enclosures is sufficient to cause the environmental conditions seen by the components under a Salem environmental profile test to be less severe than that physically attained during individual component testing.
- 2)
Determine that selected solenoid valves which are not mounted in an enclosure can withstand the Salem environmental test profile.
- 3)
Verify that the electrical connection interface of a Conax stainless steel packing gland with Grafoil grommet provides adequate electrical protection under steam, temperature and pressure conditions applicable for Salem.
Qualification Method The following steps will be taken in the course of qualifying the instrument panel and selected components:
- 1)
A thermal mode and analysis of the containment instrument panels will be prepared.
- 2)
Predict panel performance from model when SNGS-FSAR UNITS l & 2 MP79613 2 subjected to the Salem environmental test profile.
Q7.41(6)-2
- 3)
Perform environmental test on typical panel enclosures and obtain temperature gradient data -
component surface temperature, inside panel temperature, panel surface temperature and test chamber temperature.
- 4)
Verify panel model with actual test results.
- 5)
Determine component surface temperatures and correlate with previous component test data.
- 6)
Perform environmental test on solenoid valve and electrical connection interface.
- 7)
Verfiy proper operation of interface.
- 8)
Verify operation of solenoid valve and compare data with previous component test data.
The environmental test profile expected for this test is shown in figure Q7.41(6)-l.
Further test plan details will be available for your review at the PSE&G Company offices.
The instrument panel and miscellaneous component confirmatory envir~nmental tests were completed at Wyle Laboratories in Huntsville, Alabama in mid February 1979.
The following information is a summary of the test results:
SNGS-FSAR UNITS 1 & 2 MP79613 3 Q7.41(6)-3
- 1)
The instrument panel enclosures successfully passed the environmental test and maintained structural adequacy and support for enclosed transmitters and solenoids.
The panels also demonstrated its capability as a thermal barrier for the enclosed items.
- 2)
The temperatures of the interior of the panel were less than the test chamber temperature.
This demonstrates that the transmitters and solenoids enclosed in the panel experience a less severe environment than the Containment MSLB/LOCA Environment.
The peak surface temperatures of the enclosed transmitters were approximately 220-2SQOF which is considerably lower than the equipment qualification temperatures established in tests by Westing-house and Rosemount.
- 3)
ASCO solenoid valves (not enclosed in panels) successfully demonstrated operability during the SNGS-FSAR UNITS 1 & 2 MP79613 4 environmental test for isolation valve application.
The valves were the same type model as ASCO had previously tested.
This test verified their capability for the initial thermal transients (fast temperature rise times)
- Q7.41(6)-4
- 4)
The electrical interface for the transmitters, solenoids and limit switches successfully passed the test.
- 5)
The following actions will be taken based on the SNGS-FSAR UNITS 1 & 2 MP79613 5 test results:
a) Post accident instrument panels will be modified with a louvered section for pressure relief to accommodate the initial pressure transient in an MSLB/LOCA event.
b) Environmentally qualified Raychem splices instead of terminal blocks will be utilized for electrical connections inside instrument panels for post accident instrumentation.
c) The electrical interface for the Rosemount Transmitter (post accident) will be via a Conax stainless steel gland nut with grafoil grommet inserts.
d) The electrical interface for ASCO solenoid valves (post accident) will be via a Conax stainless steel gland nut with neoprene grommet inserts.
e) The electrical interface for the NAMCO EA180 limit switches will be via a potted nipple connection with an approved material such as Scotchcast *9*
Q7.41(6)-5
The concerns raised regarding the matters discussed above were addressed and proven to be acceptable at Salem during the tests that were performed.
A detailed test report will be prepared in March.
Copies will be submitted for your review.
Additional details regarding the installation of Post Accident Instrumentation inside the containment will be available for your review at the PSE&G Company offices.
SNGS-FSAR UNITS 1 & 2 MP79613 6 Q7.41(6)-6
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QUESTION 7.41(7)
For seismic qualification of safety related equipment, we require that you provide additional information on qualification reports that describe the test procedures and test results, and the manufacturer's model number, and type for each piece of equipment listed in the FSAR table Q7.18-l (see Q7.40).
ANSWER Table Q7.18-l has been revised to include the information requested on test reports and manufacturer's model numbers.
The summary forms for the Seismic Qualification Review Team have been completed and forwarded in response to Q7.40.
SNGS-FSAR UNITS 1 & 2 MP79614 1 Q7.41(7)
QUESTION 7.41(8)
Most plant areas that contain safety related equipment.
depend on the continuous operation of environmental control systems to maintain the environment in those areas within the range for which the safety related equipment installed in those areas is qualified.
It appears that there are no requirements for maintaining these environmental control systems in operation while the plant is shutdown or is not standby conditions.
(Technical Specifications only require that the control room emergency air conditioning system shall be operable).
During periods when these environmental control systems are shutdown, the safety related equipment could be exposed to environmental conditions for which it has not been qualified.
Therefore, the safety related equipment should be qualified to the extreme environmental conditions that could occur when the environmental control equipment is shutdown or these environmental controls systems should operate continuously to maintain the environmental conditions within the qualification limits of the safety related equipment.
Since you hav/fci'emonstrated that the safety related equipment installed at the Salem Unit 2 plant was qualified to the extreme environmental conditions that could occur when the environmental control equipment is shutdown, we require that an environmental monitoring system should be provided.
This environmental monitoring system should, at a minimum, meet the following requirements:
The control room should receive an alarm when the temperature range has been exceeded.
This alarm should be provided by instrumentation which should be:
(a) of high quality, (b) periodically tested to verify its functional capability by plant technical specification requirements, and (c) energized from continuous power sources.
SNGS/FSAR UNITS 1 & 2 MP79615 1 Q7.41(8)-l
The operator should have a method of maintaining a continuous record of the* temperature during the time that the environmental conditions exceed the normal limits.
We also require that you commit to provide an abnormal occurrence report to the NRC when the temperature exceeds the value for which the equipment was qualified.
In addition, if the qualification temperature is exceeded, we require that you provide the results of an analysis to demonstrate that the excess temperature has not degraded the involved Class IE equipment below an acceptable level for continued plant operation.
State your intent to meet our requirements regarding this matter.
ANSWER Refer to Question 7.32 for response.
SNGS-FSAR UNITS 1 & 2 Q7.41(8)-2
(1) Regulatory Guide 1.8, "Personnel Selection and Training," 9/75, (endorses Nl8.l-1971).
I (2) Regulatory Guide 1.30, "Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation" Electric Equipment, 8/7 2, (endorses N45. 2. 4-19 7 2; IEEE _Standard 3 3 6-1971). I (3) Regulatory Guide 1.33, "Quality Assurance Program Requirements (Operation),
11 2/78 (endorses Nl8.7-1976/ANS-3.2).
(4) Regulatory Guide 1.37, "Quality Assurance Requirements for Cleaning of Fluid Systems and Associated Compon_ents of Water-Cooled Nuclear Power Plants, 11 3/73, (endorses N45.2.l-1973).
(5) Regulatory Guide 1.38, "Quality Assurance_Requirements for Packa-ging, Shipping, Receiving, Storage and Handling of Items for Water~Cooled Nuclear Power Plants," 10/76, (endorses N45.2.2-1972),
with the following exception: identification marking on the surface of austenitic stainless steel by station personnel is permitted provided:
(a)
The halogen, sulphur and a low-melting metal content in the marking ink shall not exceed.1% by weight each.
(b)
Prior to installation of the stainless steel item, the marking ink shall be removed by cleaning with nonhalogenated solvent (acetone, alcohol or equal).
(6) Regulatory Guide 1.39, "Housekeeping Requirements for Water-Cooled Nuclear Power Plants," 3/73, (endorses N45.2.3-1973).
(7) Regulatory Guide 1.58, "Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel,"
D.2-12 J
I
( 7)
(Continued}
8/73, (endorses N45.2.6-1973).
All PSE&G personnel performing J
inspection, examination or testing shall be qualified in accordance with this Regulatory Guide, with the following exceptions:
Certification of personnel qualification will not include an expiration date and will remain in effect for the duration 0£ employment in the same classification of qualification.
Our implementation of ANSI standard
.N45.2.6, paragraph 2.2.4(5), will be by recording in documents other than the certificates, the date of certification and the bi-yearly re-evaluation of paragraph 2.2.3.
Should the re-evaluation result in a reclassification, a new certificate will be issued.
(8)
Regulatory Guide 1.64, "Quality Assurance Requirements for the Design of Nuclear Power Plants," 10/73, (endorses N45.2.ll-1974).
( 9.) Regulatory Guide 1.74, "Quality Assurance Terms and Definitions," 2/74, (endorses N45. 2.10-19 7 3).
(10) Regulatory Guide 1.88, "Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records," 10/76, (endorses N45.2.9 -
1974). The following is an I exception to Regulatory Guide 1.88: Engineering and Construction records are being duplicated via microfilm as an ongoing process.
When this activity is completed, Salem records will comply with this Regulatory Guide.
D. 2.- 13
(11)
Regulatory Guide 1.94, "Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants," 4/76 (endorses N45.2*5-1974).
Major modifications made to the Salem units will comply with Regulatory Guide 1.94.
(12) *Branch Technical Position 9.5-1, Appendix A, "Guidelines for Fire Protection for Nuclear Plants Docketed Prior to July 1, 1976," 2/77.
The Corporate QA Program shall be ~pplied to the Fire Protection Program to an extent consistent with the requirercents of Section C of Appendix A to BTP 9.5-1.
The overall QA program is described in the PSE&G Quality Assurance Manual which is prepared and maintained by 'the Quality Assurance Department.
Generally, the manual consists of:
(1)
Corporate policy statements made by responsible heads of various organizations in PSE&G involved with QA activities
- D.5.2 PROGRAM The Operational Quality Assurance (OPQA) Program requires that activities affecting nuclear safety be accomplished under suitably controlled conditions.
The program takes into account the need for procedures, special controls, cleanliness, processes, test equip-ment, tools and skills to obtain the required quality and the verifi-cation of quality by inspection, test, examination, surveillance and independent review and audit.
The safety-related activities include, but are not limited to:
designing, purchasing, fabricating, handling, shipping, storing, cleaning, erecting, installing, inspecting, testing, operating, maintaining, repairing, refueling and modifying.
These safety-related activities shall be conducted in accordance with Regulatory Guides identified in this section to include but not be limited to;
- a.
Regulatory Guide 1.8, "Personnel Selection and Training," 9/75 (endorses NlB.1-1971).
- b.
Regulatory Guide 1.30, "Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrum~ntation and 336-1971).
(endorses N45.2.4 -
1972; IEEE Standard I Electric Equipment," 8/72,
- c.
Regulatory Guide 1.33, Revision 2 (2/78), "Operational Quality Assurance Program", (endorses ANSI Nl8.7-1976/ANS-3.2).
- d.
Regulatory Guide i.37, "Quality Assurance Requirements for Clean-ing of Fluid Systems and Associated C?mponents of Water-Coolded Nuclear Power Plants," 3/73, (endorses N45.2.l-1973).
- e.
Regulatory Guide 1.38, "Quality Assurance Requirements for Pack-D.5-9 I
I
aging, Shipping, Receiving, Storage and Handling of Items for Water-Cooled Nuclear Power Plants," 10/76, (endorses N45.2.2-1972),
with the following exception:
identification marking on the surface of austenitic stainless steel by station personnel is permitted provided:
(a)
The halogen, sulphur and a low-melting metal content in the marking ink shall not exceed.1% by weight each.
(b)
Prior to installation of the stainless steel item, the marking ink shall be removed by cleaning with nonhalogenated solvent (acetone, alcohol or equal).
- f.
Regulatory Guide 1.39 "Housekeeping Requirements for Water-Cooled Nuclear Power Plants~" 3/73, (endorses N45.2.3-1973).
- g.
Regulatory Guide 1.58, "Qualification of Nuclear Power Plant Inspection, Examination, and Testing Personnel," 8/73, (endorses N45.2.6-1973).
All PSE&G personnel performing inspection, examination or testing shall be qualified in accordance with this Regulatory Guide, with the following exceptions:
Certification of personnel qualification will not include an expiration date and will remain in effect for the duration of employment in the same classification of qualifi-cation.
Our implementation of ANSI standard N45.2.6, paragraph 2.2.4(5), will.be by recording in documents other than the certificates, the date of certification and the biyearly reevaluation of paragraph 2.2.3.
Should the reevaluation result in a reclassification, a new certificate will be issued.
- h.
Regulatory Gtiide 1.64, "Quality Assurance Requirements for the Design of Nuclear Power Plants," 10/73, (endorses N45.2.ll-1974).
n.s-10
- i.
Regulatory Guide 1.74,* "Quality Assurance Terms and Definitions,
2/74, (endorses N45.2.10-1973).
- j.
Regulatory Guide 1.88, "Collection, Storage, and Maintenance of Nuclear Power Plant Quality Assurance Records," 10/76, (endorses N45.2.9-1974).
The following is an exception to Regulatory Guide 1.88: Engineering and Construction records are being duplicated via microfilm as an ongoing process.
When this activity is completed, Salem records will comply with this Regulatory Guide.
- k.
Regulatory Guide 1. 94, "Quality Assurance Requirements for Installation, Inspection, and Testing of Structural Concrete and Structural Steel During the Construction Phase of Nuclear Power Plants," 4/76 (endorses N45.2.5-1974).
Major modifications made to the Salem units will comply with Regulatory Guide 1.94.
To implement these requirements, the OPQA Program requires that safety-related activities be:
- a.
-Performed in accordance with reviewed, approved written policies and procedure throughout the life of the station.
- b.
Inspected by qualified inspectors to assure that the required quality levels have been achieved.
- c.
Subject to a formal surveillance program conducted by the SQAE, or other qualified supervisory personnel.
- d.
Subjected to systematic audits conducted by the SQAE and the QAD.
The OPQA Program shall apply to those structures, systems, and com-ponents identified in Appendix c of the Salem FSAR.
n_c;-11 I
~--~---
e where qual~ty c~n be verified by receiving inspection or installation check out, are not normally included in the Eotification Point Program.
Surveillance of suppliers/contractors during fabrication, installation, modification, repair, inspection, testing and shipment of materials, equipment and services is conducted by qualified Q,AD personnel or Q,AD agents at the supplier's/con-tractor's facility.
The SQ,AE or an EPD a.gent shall perform surveillance of activities of contractors working at the station under EPD sponsorship.
Corporate Q,AD shall perfo-rm surveillance of supplier activities at the station, except for the foregoing condition.
The surveillance is conducted in accordance with written procedures and is designed to assure conformance with procurement requirements in accordance with the safety signifi-cance of the item or service.
Consistent with the importance or complexity of the item or service, periodic evaluations of the supplier/contractor quality program is conducted.
Dependent upon the evaluation, additional audits or corrections may be required of the supplier/contractor.
Procurement of replacement parts will be by adherence to the original design criteria, where feasible (such as NSSS components in accordance with Westinghouse documentation, other code com-ponents in accordance wit_h ASME Section III 1971 and Summer 1972 Addenda or later).
This will provide the level of safety in-tended and will not result in redesign of the system.
Quality assurance requirements will be consistent with the FSAR commitments.
D.2-30
the EPD QA Program by station personnel shall be conducted by the SQAE.
The SQAE or an EPD agent shall perform surveillance of activities of contractors working at the station under EPD sponsor-ship.
Corporate QAD shall review and approve these contractors' QA Programs prior to start of work and subsequently perform audits to assure implementation.
Corporate QAD shall perform surveillance of supplier and contractor activities at the supplier facility or at the station except for the foregoing condition.
Discrepancies discovered during the conduct of surveillance and audit shall be brought to the attention of the management responsible for accomplishment of the activity.
D.2-35