B11953, Forwards Revised Responses to Sqrt Audit Questions,Per 851101 Meeting W/Nrc & S&W & Bj Youngblood Re Sser 4 (NUREG-1031)
| ML20137Q781 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 01/16/1986 |
| From: | Bishop R, Opeka J NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | Noonan V Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20137Q786 | List: |
| References | |
| RTR-NUREG-1031 B11953, NUDOCS 8602070022 | |
| Download: ML20137Q781 (10) | |
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(203) 665-s000 January 16,1936 Docket No. 50-423 B11953 Director of Nuclear Reactor Regulation Attn:
Mr. V. Noonan, Director PWR Project Directorate #5 Division of PWR Licensing - A U. S. Nuclear Regulatory Commission Washington, D. C. 20555
References:
(1)
- 3. F. Opeka letter to B. J. Youngblood, Responses to Seismic Qualification Review Team (SQRT) Audit Questions, dated October 3,1985.
(2)
- 3. F. Opeka letter to b. J. Youngblood, Revised Responses to Seismic Qualification Review Team (SQRT) Audit Questions, dated November 19,1935.
(3)
B. J. Youngblood letter to 3. F. Opeka, Supplement 4 to NUREG-1031, Millstone Nuclear Power Station, Unit No. 3, dated December 6,1985.
Dear Mr. Noonan:
Millstone Nuclear Power Station, Unit No. 3 Revised Responses to Seismic Qualification Review Team (SQRT)
Audit Questions Representatives from Northeast Nuclear Energy Company (NNECO) and Stone &
Webster met with the Staff on November 1,1985 to discuss the Staff's concerns regarding our submittal (Reference 1). In Reference (2), NNECO transmitted the revised responses to certain SQRT audit questions that remained open at the November 1,1985 ineeting. In Reference (3), the Staff concluded that the responses to all questions were acceptable except three. Attachment I provides responses to those three questions.
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. We believe that this information fully responds to the Staff's concerns, however, if you have any questions, please contact our licensing representative directly.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY et. al.
By NORTHEAST NUCLEAR ENERGY COMPANY Their Agent W. A
- 3. F.'Opeka
'~'
Senior Vice President T.
ByM'.Eshop Secretary STATE OF CONNECTICUT )
) ss. Berlin COUNTY OF HARTFORD
)
Then personally appeared before me R. W. Bishop, who being duly sworn, did state that he is Secretary of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensees herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.
//).
J Norary Putilic
"/
My Commission Expire: 1.!:r:h 31, Igg
- _ ~ - _ _ _ - -.
1 BOP-2 3 ENS *SWG A Question 1.1:
The test fixture supports the equipment horizontally; how does this simulate the inservice condition in both SS and FB directions?
f
Response
1 The test fixture has a vertical frame which supports the bus duct exiting the top of the switchgear.
This simulates the inservice condition. The bus duct is j
supported from the floor above the switchgear. Since relative displacements between floors are small (.027" N-S,.026" E-W,.0018" Vert) the test fixture accurately represents in-situ conditions.
Question 1.2:
Is there any device located on the extended structure?
Response
The 20" aluminum rear cell extension is provided to accommodate the large number of cables entering through the rear of the switchgear. No equipment is 4
mounted in the extension.
1 Questions 1.3 and 1.4:
Were relays tested with the switchgear? Justify the weight and dimensional differences between the tested and installed specimens.
Response
i The discussion in the summary report regarding the Wyle tests and how they are j
used to qualify the supplied equipment is somewhat confusing. GE has indicated that Wyle test report 43581-1 qualifies a 26" wide unit and Wyle test report l
44365-1 qualifies a 36" wide unit with bus duct. Both of these tests are stand alone qualifications. In each case relays were mounted in the switchgear and monitored in the operating mode, with the exception of two relays (PVD and 13CV). These relays are qualified generically as described in GEZ-6675 " General Electric Company, Relays and Accessories, Seismic Capabilities", to levels far in excess of those anticipated for their application at Millstone 3 (see ). Since both a 26" and a 36" unit were fully tested no significant differences exist between the tested equipment and the installed equipment.
Question 1.5:
Is the device qualification valid when tested in a different cabinet?
. Response:
As stated above both the 26" unit and the 36" unit had a full compliment of relays installed and monitored during testing.. Further, G.E. protective relays BOP-2-1
are type tested to fragility levels (CEZ-6675, see Attachment II - Test Plan).
Therefore, comparisons between the 26" unit and the 36" unit are not required to establish qualification of either unit.
Question 1.6:
Justify the qualification by similarity of the HFC relay.
Response
Both the IFC and HFC relays are generically qualified to fragility levels (GEZ-6675) which greatly exceed required qualification levels (see Attachment I).
Question 1.7:
Was there any modification of the specimen during testing?
Response
GE has reviewed the Wyle reports and found that no anomolies were reported.
Question 1.8:
Justify the RRS curve used for qualification.
Response
As discussed during the audit and in subsequent conversations with the staff, the test response spectra envelopes the amplified floor response spectra for the floor and ceiling of the switchgear area.
As discussed above, relative floor movements are insignificant.
Question 1.9:
Indicate whether special bolted mounting configuration used during testing was complied with during installation.
Response
The review of the as installed versus the tested mounting condition of the equipment was performed in detail during the site audit. This included review of vendor drawings, calculations, S&W structural drawings and S&W/ Vendor j
l correspondence. The tested mounting complied as closely as possible to the GE requirements. The installed mounting complies with GE requirements.
Question 2:
l Audit did not demonstrate whether there was sufficient clearance between the top point of the switchgear and bus duct to avoid interference.
Response
f l
This concern, which was raised during the site audit, motivated the installation of a modified bus duct to provide adequate clearance. This work has been completed on E&DCR N-EC-00846 which has been closed out.
BO P-2-2 i
I
NSSS-2 3S!H*P Question 1 Torsional frequency of assembly must be computed and compared to motor's operational speed.
Response
This calculation has been performed with the following results:
Operating speed 3570 rpm ist torsional critical speed 4505 rpm This is 26% above operating speed and 62% below twice the motor electrical power line frequency of 7200 rpm.
NSSS-2-1
NSSS-6-3R PS
- SWG 1 Question 2:
Proximity of the installed switchgear to an adjacent non-Category I cabinet and the possible dynamic interaction between the two cabinets were not addressed in the qualification documents.
Response
A response to the above question was provided to the NRC in a letter dated November 19,1985. (Bil878).
NRC Position Regarding NNECO Response:
The applicant must confirm adequacy of the modification to preclude dynamic interaction.
Revised Response: (1/86)
The evaluation of mounting a Class IE reactor trip switchgear to a non-Class IE rod drive control cabinet (M-G) was completed. The evaluation indicated that the seismic qualification of the Class IE reactor trip switchgear (RTS) would be retained if the RTS unit was connected to the adjacent rod drive control cabinet.
The above conclusion is based on the following study:
1.
Review of the structural integrity of the rod drive control cabinet (Reference 1 and 3) to document its adequacy.
The structural drawings and photographs of the rod drive control cabinet were reviewed. It was concluded that the construction of this cabinet is similar to that of the trip switchgear. It was identified that this cabinet has most of its internal weight located toward the bottom of the cabinet.
Therefore based on this review of the cabinet details and its contents, it was concluded that the equipment has sufficient structural integrity to.
survive the seismic event.
2.
Evaluation of the seismic dynamic response characteristics of the combined rod drive /switchgear assembly (Reference 3).
Calculations were made to estimate the effect of coupling the two units on the natural frequencies of the combined system. It was determined by analysis that the shif t in front-to-back frequency would be less than 10 percent when the two cabinets are connected. Due to the broadness of the RRS used to qualify the RTS, this change is insignificant. Likewise the combined side-to-side frequencies were estimated and determined that connecting the cabinet tops will increase the side-to-side natural frequencies of the cabinet assembly.
3.
Review of the results of the generic seismic tests performed on the reactor trip switchgear (Reference 2).
NSSS-6-1
The results of the generic seismic test were reviewed to estimate the generic peak interaction force that could be developed between the two cabinets.
4.
Design of an interface connection strong enough to transmit the largest likely interaction forces between the two cabinets (Reference 3 and 4).
Based on the loads defined in Step 3, a generic bracket design was developed.
This generic design was then adjusted to fit the as-built mounting conditions present at Millstone Unit No. 3.
References:
1.
Calculation, " Evaluation of RD Control Cabinet".
2.
Report #SOTAR-80-3, " Seismic Qualification of Reactor Trip Switchgear Assembly with Type DS Model 416 Breakers".
3.
Calculation, "RTS/M-G Set Cabinet Interface".
4.
Calculation,"MG/RTS Set".
NSSS-6-2
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APPENDIX 0 g
- q. ; ;* ~ a-r..l L* l.,;. l. 7.a ![
POWER SYSTEMS
- 7 6
t MANAGEMENT B USIN ESS I
GENERAL ELECTAiO COMDANY 205 GAEAT VALLEY PAR *(WAY, MALVERN. PA 19355-0715 i
DEPARTMENT Pnone (215) 251-7000 l
1E82352 Engineering Sectior.
August 31, 1982
Subject:
NUCLEAR CERTIFICATION OF RELAYS, SWITCHES AND ACCESSORY DEVICES
Reference:
REQUISITION NO. 297-84762 FOR: MILLSTONE SUW.ARY NO.
NUCLEAR NO. 8236-46C T0:
Mr. W. P. Flowers, MD 706 Development and Design Engineering General Electric Company - MVSBS P. O. Box 4SS Burlington, Iowa 52601 The attached list shows the seismic capability and/or qualified life of the listed models. This determination has been accomplished either through testing per IEEE C37.98-1978 (Seismic Testing of Relays) and the General Electric Co. interpretation of IEEE Standard 323-1974 (Qualifying Class 1E Equipment for Nuclear Power Generating Statior.s) or by comparative analysis to tested devices.
The attached list indicates how the capability of each device was determined.
A copy of the generic test plan for qualifying class IE devices is attached.
A condensed description of the procedure used in the testing program is shown on the attached GEZ-6675 for Seismic Testing and GEZ-7029 for Qualification Testing.
The models on the attached list are qualified for the following service conditions:
1.
Altitude - 5,000 ft. maximum 5.
Radiation - 1 x 104 rads 2.
Current or voltage - rated (total integrated dose) 3.
Frequency - rated
- 6.
Relative humidity - 60% average 4.
Temperature - 300C average with 10-90% excursions with 100C-400C excursions Detailed test data for each test program are available on a mutually agreed schedule for audit at the Power Systems Management Business Department of the General Electric Co. There will be a charge for such audits, which is commensurate with the time and expense incurred by General Electric for such audits. The location of the Engineering Of fice for these data is 6901 Elrrrsood Avenue, Philadelphia, PA 19142.
In order to identify the requisition and the customer order number for which the audit will be performed, a copy of this paper must accompany all such requests for audit.
Very truly yours, V. F. Scamman Senior Design Engineer Nuclear Qualification 9
+ ~
a --
f e
1E82362 QUALIFICATION CAPABILITIES - CLASS 1E DEVICES Page 1 of 7 SEISMIC FRAGILITY LEVEL IN g's ZPA NONOPERATE MODE OPERATE MODE
- 1) QUAL.
QUALIFIED LIFE TRANSITION TES1 MARK METHOD
- 1) METHOD TIME IN SECONDS FILE NO.
A YRS 1
A N0HAAL SEISHIC REF.
NO.
I35G 01 10AX008G10 Seismic capability is 4 g ZPA X 2)10 X
- g SCM-001 l37Z 02 006353570G012 Seismic capability is 6 9 ZPA X
41 X
SB12-001 9A7 03 16589AB302STS2P Seismic capability is 4 g ZPA X 2)10 X
SB9-001 l 909 04 16SB9KB2A10LSM2P Seismic capability is 4 g ZPA X 2)10 X
(,,}9El 05 16SB98F3A52STT2P Seismic capability is 4 g ZPA X 2)10 X
509-001 9E2 06 16SB9FB3A52STT2P Seismic capability is 4 g ZPA X 2)10 X
SB9-001
'fby/llj9G2 07 16SB9CB3655STS2P Seismic capability is 4 g ZPA X 2)10 X
549-001
}l9G6 08 16589RB2A06LSM2P Seismic capability is 4 g ZPA X 2)10 X
589-001 i
9A8 09 16SB9FD333SSL2K Seismic capability is 4 g ZPA X 2)10 X
509-001 9A9 10 16589F0509S5V2K Seismic capability is 4 g ZPA X 2)10 X
589-001 i
L,'
981 11 16SB9BB408SSM2K Seismic capability is 4 g ZPA X 2)10 X
SB9-001 932 12 \\16SB9CB204SDM2Y Seismic capability is 4 g ZPA X 2)10 X
589-001 6) 82 13 11686708G043td & Seismic capability is 6 g ZPA X
41 X
Ell 6-001
.x 781 14 EB25A04BC Seismic capability is 6 g ZPA X
41 X
ED25-001
,'O 78N 15 EB25A06BC Seismic capability is 6 g ZPA X
41 X
lB25-001 N
(78R 16(EB25A12BC Seismic capability is 6 g ZPA X
41 X
ERPS-001 t
NOTES:
- 1) T - Qualified by Test; A - Qualified by Analysis (comparison to tested device)
- 2) Qualification tests or a similar device are being continued to extend Qualified Life beyond ten years.
,7
. F.
camman (MD1104)
Senior Design Engineer Nuclear Qualification Regn. No.
297-84/62 Power Systems Management Business Department General Electric Company Nuclear No.
82_36-460 205 Great Valley Parkway Malvern, PA 19355-0715 DATE:
_ A_u g u s_L_31_,, 1982
1E82362 QUALIFICATION CAPABILITIES - CLASS lE DEVICES Page 2 nr 2 SEISMIC FRAGILITY LEVEL IN g's ZPA NONOPERATE MODE OPERATE MODE
- 1) QUAL.
QUALIFIED LIFE TRANSIT!Ce TEST MARK METHOD.
- 1) METl!0D TIME IN SECONDS FILE NO.
A YRS T
A NORMAL SEl3RIC REF. NO.
I1N9 01 12HEA51C239X2 4 4
4 4
X 2)10 X
HEA-001
! 987 02 3)12HFA151A2F 3
2 6
6 X
41 X
.086 PU.055.118 HfA-002 (D.C.)
.41 X
.020
.011.034 firC-001
.037 DO.028.042 b) 9A6 03 12HFC2181A 3.5 6
X
}k 9E3 04 12HGA111J2 3.5
.0.5 1.75 4 X
41 X
.075 PU.060.081 IIGA-002 (D.C.)
.035 00.032.049 7
,9C2 05 12IFC51ADIA 3
5 X
41 X
.72
.59-1.03 IFC-001 9C6 06 12IFC66K1A 3
4 5
3.5 X
41 X
6.87 5.84-7.79 IFC-007 9J7 07 121JCV51821A 3
3.5 4
4 X4)41 X 10C.67
.63.86 IJCV-001 IOC.014
.010.020
. 9AS 08 12NGV13B21A 6
6 6
6 X
41 X
.026
.018.039 NGV-001 (904 09 12PVD21BIA.
3 2
X 41 X
.051
.033.077 PVD-001 9C9 10 12 SAM 11B22A 6
4 6
6 X
41 X
.028
.018.035 SAN-007
- 1) T - Qualified b' Test; A - Qualified by Analysis (comparison to tested device)
NOTES:
y
- 2) Qualification tests on a similar device are being continued to extend Qualified Life beyond ten years.
- 3) Seismic capability values apply to contact codes 33, 42, 51 and 60 for three or less N.C. (nonnally closed) contacts.
N.C. contact seismic capability may be reduced for contact codes 06,15 and 24 with four or more N.C. contacts.
- 4) Continuous current limits for 41 years Qualified Life for 2-8 and 4-16 amp Instantaneous Units are 7.75 and 5.5 amps respectively.
fo wg
~
V. F. Scamman (MD1104) i Senior Des ~ign Engineer Nuclear Qualification Regn. No.
297-84762 Power Systems Management Business Department General Electric Company Nuclear No.
8236-460 205 Great Valley Parkway Malvern, PA 19355-0715 DATE:
_Aygust 31 1992 1
1