HL-5227, Forwards Response to RAI Re Unresolved Safety Issue A-46
| ML20117C160 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 08/23/1996 |
| From: | Beckham J GEORGIA POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR HL-5227, NUDOCS 9608270301 | |
| Download: ML20117C160 (31) | |
Text
a Georgia Power Company 40 inverness Center Parkway Pest Office Box 1295 Birrdogham Alabama 35201 TetIphone d'oS 877-7279 hce'hresden f0 clear GeorgiaPower Hatch Project te soveun recfnc sprcn August 23, 1996
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Docket Nos. 50-321 HL-5227 50-366 U.S. Nuclear Regulatory Commission A'ITN: Document Control Desk Washington, D.C. 20555 Edwin I. Hatch Nuclear Plant Response to Request for Additional Information:
Unresolved Safety Issue A-46 Gentlemen:
j Per Nuclear Regulatory Commission (NRC) letter dated June 27,1996, Georgia Power i
Company (GPC) submits the enclosed response to the request for additional information regarding Unresolved Safety Issue A-46. The enclosure lists the NRC Requests followed by the GPC response.
Should you have any questions in this regard, please contact this office.
Sincerely, f
f J. T. Beckham, Jr.
i DLM/eb
Enclosure:
. Response to Request for AdditionalInformation:
- Unresolved Safety Issue A-46 cc:
Georgio Power Company
}
Mr. H. L. Sumner, Nuclear Plant General Manager j/
NORMS US Nuclear Regulatory Commission. Washington. D.C.
.:.<0098 Mr. K. Jahdour, Licensing Project Manager - Hatch U.S Nuclear Regulatory Commission. Region II
. Mr. S. D. Ebneter, Regional Administrator
- Mr. B. L. Holbrook, Senior Resident Inspector - Hatch 9608270301 960823 PDR ADOCK 05000321 P
j e,
i Edwin I. Hatch Nuclear Plant Response to Request for Additional Information:
l Unresolved Safety Issue (USI) A-46 1.
NRC Request:
In Appendix G of Reference 1, three different in-structure response spectra (IRS) curves for three different soil conditions Gower, intermediate, and upper soil i
modules), were included at 87 ft.,130 ft. and 158 ft. elevations. Indicate which in-structure response spectrum was used to calculate seismic demand for the safe l
shutdown equipment at these elevations.
GPC Resnonse:
)
The IRS used to calculate seismic demand are discussed in section 4.2.1.2 of the Unit 1 USI A-46 Summary Report, section 4.2.1.3 of the Unit 2 USI A-46 Summary Report (Reference 1), and sections 5.3,5.4, and Appendix B of l
EPRI NP-7212-SL (Reference 2). The seismic demand is based on the envelope of the IRS curves for the three different soil conditions (lower, intermediate, and upper soil modulus profiles) at a given location and direction. Broadening is also included with the enveloping to define seismic demand. For equipment supported at a given elevation,'the seismic demand for a given direction was calulated based on the broadened envelope of the three soil case IRS at that elevation and direction.
' 2.
NRC Regnect:
The IRS curves shown in Appendix G are not broadened by il5 percent. Submit i
copies of the actually broadened IRS curves that were used for elevations 203 ft.,
228 ft., and 280 ft. in the reactor building and elevations 130 ft. and 150 ft. in the diesel generator building.
4 GPC Resnonse:
J As stated in section 4.2.1.2 and Appendix G of the Unit 1 USI A-46 Summary Repod, and section 4.2.1.3 and Appendix G of the Unit 2 USI A-46 Summary Report (Reference 1), the IRS curves shown in Appendix G are the raw seismic margin eanhquake (SME) 5 percent damped IRS. The three raw IRS associated with the lower bound, the intermediate, and the upper bound soil modulus profiles for a given location, direction, and damping value were plotted together as part of the original Plant Hatch Seismic Margin Assessment (SMA) (Reference 2). Rules 4
to broaden the appropriate raw SME IRS are discussed in the Unit I and Unit 2 5
HL-5227 E-1
i Response to Request for Additional Information:
U,nresolved Safety Issue (USI) A-46 USI A-46 Summary Reports in the sections referenced above. The rules for i
broadening and enveloping the three soil case IRS were used to defme the seismic demand; however, no formal plotting of the broadened envelope IRS was done. A single peak spectral acceleration value for a given direction was typically used in the evaluation; therefore, plotting the broadened envelope spectra was unnecessary.
With the raw spectra, and the rules for broadening and enveloping, the seismic capability engineer was fully able to determine the seismic demand for any item of equipment under evaluation. Therefore, the IRS curves shown in Appendix G represent the curves used for the subject evaluations.
-3.
NRC Request:
Two dimensional (2-D) stick models are shown in Figures G-2 and G-3 for the reactor and the diesel generator buildings, respectively. However, it is not clear whether GPC used a 2-D or 3-D stick model for generating the IRS. If a 2-D model was used, specify which two directions were considered in the analysis. If a 3-D model was used, provide the 3-D stick model used for the analysis.
GPC Resnonse:
1 A detailed discussion of the structural models used in the A-46 evaluation was referenced in the GPC 120 day response to Generic Letter 87-02, Supplement 1.
Specifically, the structural models are discussed in Section 5.2 and Appendix B of Reference 2. A brief summary is provided below.
The reactor building and the diesel generator (DG) building seismic building models are 2-D; however, there is a separate North-South / vertical and an East-West / vertical 2-D seismic model for each of these two buildings. The figure of the seismic building model for the N-S/ vertical direction and the E-W/ vertical direction are similar; therefore, the N-S and E-W directions were not included in the title of Figures G-1 and G-3. The appropriate N-S/ vertical 2-D seismic building model was used to calculate the N-S SME IRS, and the appropriate E-W/ vertical 2-D seismic building model was used to calculate the E-W SME IRS for each of these buildings.
4.
NRC Reauest:
With respect to the stick model shown in Figure G-3, provide elevations and magnitudes of Mass 1 and Mass 2. In addition, provide the types of boundary conditions applied at the masses, identify the mass at which the input ground accelerations were applied, and the stiffness of the element between the two masses.
Discuss the factors that contribute to the almost identical IRS and no relative displacements between Masses 1 and 2. Also, provide the absolute displacement of the two masses.
HL-5227 E-2
1 Response to Request for Additional Information:
U,nresalved Safety Issue (USI) A-46 I
i i
GPC Response:
1 2
i Mass point 1 is defined at el 150 ft and has a total mass of 372.59 k-sec /ft or 1
11,997.4 kips. Mass point 1 is the roof of the DG building. Mass point 2 is defined
. at el 130 ft and has a mass of 633.27 k-sec /ft or 20,391.2 kips. Mass point 2 is the 2
5-ft-thick base slab of the DG building which is founded at grade. Member 1 connects mass points 1 and 2. The N-S member properties are: 1) axial area =
2 2
d 2438 f1, 2) shear area = 1394.5 f1,and 3) area moment ofinertia = 10,109,311 ft.
2 2
The E-W member properties are: 1) axial area = 2438 ft, 2) shear area = 1066 ft,
and 3) area moment ofinertia = 3,312,928 ft'. The modulus of elasticity is 526,000 ksf.
A discussion of the DG building soil-structure interaction (SSI) analysis is provided in Appendix B of Reference 2 which includes the following statement on page B-4:
"The bottom of the basemat of the diesel generator building is located at el.125 ft,5 ft below grade. Because of the shallow embedment, impedance functions were calculated for a flat foundation on a truncated soil column extending to a maximum elevation of 125 ft and the foundation input motions were taken to be equal to the free-field motions; i.e., scattering was unity."
Table B-3 of Reference 2 lists the fixed base natural frequencies of the DG building. The N-S fundamental natural frequency is 31.43 Hz; the E-W fundamental natural frequency is 27.34 Hz; and the vertical fundamental natural frequency is 66.02 Hz. This information demonstrates that the DG building is a relatively rigid one-story concrete box structure set on soil.
The DG building has a very low aspect ratio of height to plan dimension of the base slab which is a 5-ft-thick mat foundation. The height from the top
. of Jhe base slab to the top of the roof slab is only 20 ft; whereas, the length i
of the DG building foundation in the N-S direction is approximately 191 ft, and the length in the E-W direction is approximately 104 ft. Based on these facts, minimal rocking and small relative displacements between the base slab and the roof would be expected, since most of the displacement would be pure translation of the rigid box structure.
i Plant Hatch was the first plant to conduct a combined SMA and USI A-46 evaluation.- The Plant Hatch SMA was an EPRI pilot plant program for a BWR and a soil site. The NRC was actively involved in this study, which is documented in Reference 2, including reviews by the NRC Seismic Design Margins Working Group, NRC staff, an NRC Peer Review Group (PRG) composed of five industry experts, and an NRC consultant involved in the i
i HL-5227 E-3 I
i
I Response to Request for Additional Information-U.nr,esolved Safety Issue (USI) A-46 i
USI A-46 programs. References 3,4,5, and 6 document the review reports from the NRC PRG and the Seismic Design Margins Working Group. The I
NRC included the development of the SME IRS in their review. The NRC requested Dr. M. P. Bohn, a member of the NRC PRG, to conduct an independent review of the SSI analysis performed for the Hatch SMA. By letter to the NRC dated July 5,1991, Dr. Bohn transmitted the results of the Hatch study (Reference 4). In the report, Dr. Bohn makes the following statement regarding the DG building SSI results:
"... there is relative little amplification of the structure over the ground motion input and thus the building is behaving in a relatively rigid fashion, as expected. Furthermore, the area of the foundation in comparison to the height of the structure results in motion being primarily horizontal with little rocking components... Thus, it can be concluded that the spectra j
generated for the Hatch Diesel Generator building are realistic l
and can be reproduced by independent methods."
As described previously, the absolute displacement of these two masses was considered insignificant and consequently, was not calculated.
4 5.
NRC Request:
i For plant structures containing equipment in the USI A-46 scope:
j 1
l 1
a.
Identify structures which have licensing-basis floor response spectra
}
(5% critical damping) for elevations within 40-feet above the effective grade,
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which are higher in amplitude than 1.5 times the SQUG Bounding Spectrum.
l l
b.
Provide the response spectra designated according to height above the effective grade identified in Item a. above and a comparison to 1.5 times the j
Bounding Spectrum.
I c.
With respect to the comparison of equipment seismic capacity to seismic I
demand, indicate which method (Method A or Method B in Table 4-1 of j
GIP-2) was used to address the seismic adequacy of equipment installed on those floors as identified in Item a. above.
GPC Resonnse:
Request No. 5 is related to the use of 1.5 times the plant safe shutdown earthquake (SSE) ground response spectra as a realistic estimate of seismic demand under certain limited conditions specified in the Seismic Qualification Utility Group HL-5227 -
E-4
Response to Request for Additional Information:
Unresolved Safety Issue (USI) A-46 (SQUG) Generic Implementation Procedures (GIP). The use of 1.5 times the plant SSE ground response spectra was previously reviewed and approved by the NRC Staff. It is Georgia Power Company's (GPC) understanding that the NRC Staff and representatives of SQUG arejointly seeking generic resolution of this issue.
Accordingly, a response to this request is being deferred pending resolution. It is GPC's position that the GIP was approved by the NRC Staffin Supplemental Safety Evaluation Report No. 2, dated May 22,1992, as an acceptable method of demonstrating the seismic adequacy of equipment within its scope. The GIP methodology differs from the methodology used in original plant licensing during the 1970s in substantial and fundamental respects. Accordingly, it is impossible to meaningfully compare isolated aspects of the two methodologies including their relative conservatisms. Any such comparison must be made at the program level to evaluate compliance with appropriate NRC regulations concerning seismic adequacy.
6.
NRC Request:
Provide physical descriptions (i.e., dimension, weight, support system, anchorage, embedment, etc.) and GPC's evaluation methodology used for the following tanks and heat exchangers:
(1) Ell-B001 A-RHR Heat Exchanger A, (2) R43-A001 A-Fuel Day Tank 1 A, (3) R43-A002A-Fuel Storage Tank 1 A, (4) R43-A007A-Air Receiver and (5) T48-A001-Unit 1 Nitrogen Storage Tank.
It appears from Appendix I of Reference 1 that no tanks and heat exchangers evaluated were located above elevation 130-ft Indicate whether any tanks and heat
~
exchangers, in the safe shutdown path, are located above elevation 130-ft.
GPC Response:
As described in Appendix I of Reference 1, no tanks or heat exchangers on the Safe Shutdown Equipment List (SSEL) are located above el 130 ft.
E11-B001 A:
Residual heat removal (RHR) heat exchanger A is a vertical heat exchanger bolted 1
to steel floor beams at reactor building el 107 ft with horizontal stabilizing rods at el 118 ft-101/2 in. The heat exchanger has a height of 25 ft-51/2 in, an outside diameter of 54-3/4 in, and a total weight of 63,000 lb A seismic analysis of the heat exchanger was performed for A-46 using the Hatch SME IRS. The seismic HL-5227 E-5
i 4
i Response to Request for Additional Information:
{
Unresolved Safety Issue (USI) A-46 loads were combined with other loads such as the nozzle and dead loads. The j
results of this analysis demonstrated that the anchorage and support members are acceptable.
)
R43-A001 A-1 h
Fuel oil day tank 1 A, located at el 130 ft in the DG building, is a horizontal tank j_.
. supported by two saddles 40 in apart. The saddles are mounted directly to the concrete floor with 1 in diameter cast-in-place bolts. The anchor bolts are sleeved anchors with a 4 in x 4 in plate at the end and an embedment of 2 ft-6 in. The tank is 78 in long, and has an outside diameter of 66 in and a total weight of 9400 lb.
The USI A-46 seismic evaluation of the tank anchorage was based on a pseudo-static approach using the peak spectral acceleration values of all three SME earthquake components. A review of the original stress report for the tank saddles indicated adequate margin exists to envelope the SME seismic demand.
R43-A002A:
Fuel oil storage tank 1 A is a buried tank in the plant yard below grade el 130 ft The tank is 48 ft long, and has an outside diameter of 144 in and a total weight of 347,800 lb. A drawing review revealed the tank is placed in well compacted engineered backfill to properly cradle and cover the tank. The SMA methodology (Reference 7) states that buried tanks can be screened out at SME levels less than i
0.5g peak ground acceleration (pga). The Hatch SME is a 0.3g pga; therefore, the tank was screened out. The Hatch SMA soils evaluations show no significant relative soil displacement for the area where the tank is buried; therefore, there are i
no concerns with piping connections.
t R43-A007A:
The DG air receiver tank, located at el 130 ft in the DG building, is a small vertical i
tank with a full 360 ring stand at the base. The tank is 109 in high, and has an outside diameter of 30 in, and a total weight of 1482 lb. Four angle feet are
-l attached at the base of the tank as part of the ring stand. The angle feet are anchored to the base slab with four 3/4 in diameter, cast-in-place J bolts which have a 90 bend and an embedment of at least 18 in. The A-46 evaluation of the anchor bolts, angle feet, and vertical tank ring stand was based on a pseudo-static approach using the peak spectral acceleration values of all three SME earthquake components.
HL-5227 E-6 i
.. _ =
1.
Response to Request for Additional Information:
Unresolved Safety Issue (USI) A-46 i
T48-A001A:
f j
The Unit I nitrogen storage tank, located in the plant yard at el 130 A, is a horizontal tank supported by two saddles 32 ft apart. The tank foundation is a j
reinforced concrete mat foundation at grade with two 2 ft x 10 ft x 1 ft high concrete piers fully tied to the mat with reinforcing steel. The tank saddles are bolted to a I in thick steel sole plate with four 1-3/8 in diameter bolts per saddle. The sole
}
plate is anchored to the concrete pier with four 1-3/8 in diameter cast-in-place J bolts with an embedment of 34 in. The tank is 41 ft-1-1/2 in long and has an outside diameter of 11 ft and a total weight of 116,800 lb. The A-46 seismic evaluation of the tank anchorage and sole plate was based on a pseudo-static approach using the peak spectral acceleration values of all three SME earthquake components. The original stress report for the tank saddles was reviewed and scaled up using the same Hatch peak SME spectral acceleration values used to evaluate the anchorage and sole plate.
7.
NRC Reget:
GPC reported a few outliers of the conduit and cable raceway supports and their resolution in Appendices J and K. Provide detailed information (i.e., physical descriptions including dimension, material properties, etc.) and GPC's analytical evaluations for the resolution of nonconformances in conduit and cable raceway supports:
Anpendix J (1) 1D-63,1E-9A (2) - 1W-1 A,2U-6A, and (3) Multiple Cable Spreading Room Supports (Drawing No. H-13217).
Apnendix K (1) 1D-28 (Drawing No. H-13203),1D-68 (Drawing No. H-13204),
(2) Expansion anchors (Drawing Nos. H-13216, H-13217 and H-23208)
(3) Cable tray support 1 Y-5A (Drawing No. H-17262)
(4) Supports IU-10ED (Drawing No. H-17355) and 2U-6A (Drawing No.
H-27278).
GPC Response:
The following is a summary description of the supports and the requested evaluations. The analytical evaluations incorporate thejudgment and expertise of the Seismic Review Team (SRT) members. In lieu of detailed information and to HL-5227
.E-7
Response to Request for Additional Information:
Unresolved Safety Issue (USI) A-46 provide for the efficient use of both GPC and NRC staff resources, GPC would be pleased to provide SRT member support for an NRC review of the detailed A-46 cable and conduit support evaluation:, available at Southern Nuclear Operating Company (SNC) Corporate Offices.
Apnendix J:
Supnort 1D-63:
Raceway support 1D-63, located at control building el 112 ft-0 in, was selected by
. the SRT as one of the representative sample supports to be included in the GIP limited analytical review. An as-built sketch of the support, which provides a physical description including dimensions and materials gathered from the walkdown, is included in Attachment A. Following the limited analytical guidelines given in GIP section 8.3, the potential non-ductile failure of the anchor bolts and other support components was evaluated using the dead load check, the vertical load check, and the lateral load check. The results of the evaluation demonstrated that ID-63 meets all the requirements of GIP section 8.3 with no
. outliers.
Supoort 1E-9A:
Raceway support 1E-9A is located at control building el 130 ft-0 in. This support, along with IE-9B and IE-9C, was chosen for limited analytical review by the SRT based on the discovery of a small crack in the concrete ceiling. The crack passed through the shear plane of two of the concrete expansion anchor bolts. An as-built sketch of the support, which provides a physical description including dimensions and materials gathered from the walkdown, is included in Attachment B. Following the limited analytical guidelines given in GIP section 8.3, the potential non-ductile failure of the anchor bolts and other support components was evaluated using the dead load check, the vertical load check, and the lateral load check. The load carrying capacity of the expansion anchors was calculated taking into account the effect of the small concrete crack on the bolt capacity. Based on the results of this evaluation, the demand load calculated for the anchor bolts was less than the reduced bolt capacity, considering the presence of the small concrete crack.
Therefore, raceway support IE-9A is acceptable for A-46.
Supoort 1W-1 A:
i
' Raceway support 1 W-1 A, located at Unit I reactor building el 202 ft-0 in, was selected by the SRT as one of the representative sample supports to be included in the GIP limited analytical review. An as-built sketch of the support, which provides a physical description including dimensions and materials gathered from the j
HL-5227 E-8
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Rcsponse to Requ:st for AdditionalInformation:
Unresolved Safety Issue (USI) A-46 walkdown, is included in Attachment C. Following the limited analytical guidelines given in GIP section 8.3, the potential non-ductile failure of the anchor bolts and other support components was evaluated using the dead load check, the vertical load check, and the lateral load check. The results of the evaluation demonstrated that 1W-1 A meets all the requirements of GIP section 8.3 with no outliers.
Supnort 2U-6A:
Raceway support 2U-6A is located at Unit 2 reactor building el 130 ft-0 in. From the field walkdown, the SRT discovered that a 14 in diameter process pipe was attached to the cable tray support member and, therefore, labeled the support an outlier for further evaluation. An as-built sketch of the support, which provides a physical description including dimensions and materials gathered from the walkdown, is included in Attachment D. Following the limited analytical review guidelines given in GIP section 8.3, the potential non-ductile failure of the anchor bolts and other support components was evaluated using the dead load check, the vertical load check, and the lateral load check. The results of the evaluation demonstrated that the demand load calculated for the anchor bolts exceeded the bolt capacity when the attached pipe loads were considered. This outlier was resolved by disconnecting the pipe from support 2U-6A, thereby reducing the anchorage loads to an acceptable level.
Multiple cable spreading room supoorts:
Please reference the answer to request for additional information (RAI) No. 8 below.
Appendix K:
Support 1D-28:
Raceway support 1D-28 is located at control building el 112 ft-0 in. From the field walkdown, the SRT discovered that one of the four anchor bolts was not installed and, therefore, labeled the support an outlier for further evaluation. An as-built sketch of the support, which provid s a physical description including dimensions and materials gathered from the walkdown,is included in Attachment E. Following the limited analytical review guidelines given in GIP section 8.3, the potential non-ductile failure of the three installed anchor bolts and other support components was evalunted using the dead load check, the vertical load check, and the lateral load check. The results of the evaluation demonstrated that the demand load calculated I
using only three anchor bolts did not exceed the anchor bolt capacity. Therefore, raceway support 1D-28 is acceptable for A-46.
HL-5227 E-9
Response to Request for Additional Information:
i Unresolved Safety Issue (USI) A-46 4
4 1
Supoort ID-68:
Raceway support ID-68, located at control building el 112 ft-0 in, was selected by the SRT as one of the representative sample supports to be included in the GIP limited analytical review. An as-built sketch of the support, which provides a physical description including dimensions and materials gathered from the walkdown, is included in Attachment F. Following the limited analytical guidelines given in GIP section 8.3, the potential non-ductile failure of the anchor bolts was evaluated using the dead load check, the vertical load check, and the lateral load check. The results of the evaluation demonstrated that the demand load calculated for the anchor bolts from the lateral load check exceeded the bolt capacity. This outlier was resolved by installing a knee brace (shown in Attachment G) to reduce the loads on the existing anchors.
Supoort lY-SA:
i Raceway support lY-5A, located at control building el 112 ft-0 in, was selected by the SRT as one of the representative sample supports to be included in the GIP limited analytical review. An as-built sketch of the support, which provides a physical description including dimensions and materials gathered from the walkdown, is included in Attachment H. Following the limited analytical guidelines given in GIP section 8.3, the potential non-ductile failure of the anchor bolts and other support components was evaluated using 3: dead load check, the j
vertical load check, and the lateral load check. The resd J :he evdation i
demonstrated that lY-5A meets all the requirements of GlP section 8.3 with no j
outliers.
Sunnort IU-10ED-Raceway support IU-10ED, located at Unit I reactor building el 130 ft-0 in, was selected by the SRT as one of the representative sample supports to be included in the GlP limited analytical review. An as-built sketch of the support, which provides 4
a physical description including dimensions and materials gathered from the i
walkdown, is included in Attachment 1. Following the limited analytical guidelines given in GIP section 8.3, the potential non-ductile failure of the anchor bolts was t
evaluated using the dead load check, the vertical load check, and the lateral load j
check. The results of the evaluation demonstrated that the demand load calculated for the anchor bolts from the lateral load check exceeded the bolt capacity. This outlier was resolved by replacing the existing vertical strut member and redesigning the support anchorage as shown in Attachment J.
!'j HL-5227 E-10 t
- Response to Request for Additional Information:
Unr,esolved Safety Issue (USI) A-46 8.
NRC Reauest:
GPC indicated that Dr. R. Kennedy of Structural Mechanics Consulting performed a third-party audit of the Hatch Unit 1 USI A-46 program. In his review, he expressed some concern about the method used to determine the capacity of the cable spreading room cable tray support bolts (e.g., Wej-It expansion anchor bolt).
Accordingly, GPC revised the calculations and Dr. K.ennedy later reviewed and accepted the calculation. GPC is requested to submit the following information for staff's further review:
(1) GPC's original calculations for the capacity of the cable spreading room cable tray support bolt, (2) Dr. Kennedy's reported concern, and (3) GPC's revised calculations for the supports.
GPC Resnonse:
The following is a summary description of the requested calculations. The analytical evaluations and calculations incorporate the judgment and expertise of the SRT members. In lieu of detailed information and to provide for the efficient use of both GPC and NRC staff resources, GPC would be pleased to provide SRT member support for an NRC review of the detailed A-46 evaluations available at SNC Corporate offices.
Dr. Kennedy's reported concern is documented in the Units 1 and 2 A-46 Summary Reports as Appendix M (Reference 1.) Dr. Kennedy's initial third party review of the cable spreading room cable tray support evaluation produced two changes to the evaluation. A description of GPC's calculations and their revision, based on Dr. Kennedy's comments, is provided below.
A. The initial cable spreading room evaluation determined the pullout capacity of j
Wej-It wedge expansion anchors by applying a factor of safety of 3 to the mean ultimate pullout capacity determined from the SQUG/EPRI e_xpansion anchor test program. The evaluation was performed shortly after the test program was completed, and the GIP capacity reduction factors had not been revised to incorporate this data. SQUG later studied the data and recommended a capacity reduction factor of 0.50 for Wej-It wedge expansion anchors. Dr.
Kennedy suggested that the cable spreading room evaluation be revised using the pullout capacity recommended in the GIP with the 0.50 capacity reduction factor. This change was made and approved by Dr. Kennedy in his follow-up review, as documented in Appendix M of Reference 1.
i HL-5227 E-11
Response to Request for Additional Information:
Unresolved Safety Issue (USI) A-46 B.-
Dr. Kennedy's second comment concerned the method used for determining pullout capacity for expansion anchors with less than the recommended minimum embedment. The initial evaluation determined the pullout capacity as the smallest of either 1) the pullout capacity based on shear cone failure per l
Reference 8 or 2) the mean ultimate pullout capacity determined from the SQUG/EPRI expansion anchor test program with a factor of safety of 3 applied. Dr. Kennedy agreed with thh approach but suggested the method described in Reference 9, which had been recently released at that time and undergone intense scrutiny by SQUG and other industry leaders, might present a more acceptable approach. The evaluation was subsequently revised to i
reflect use of the method described in Reference 9 and was approved by Dr.
Kennedy in his follow-up review (Reference 1).
j 9.
NRC Reauest:
Section 2.3.1," Primary Shutdown Path" states in part that, "If the DGs do not auto-start as a result of relay chatter, the emergency operating procedures (EOPs) direct an operator to be dispatched to the diesel building to assess conditions. Indications on the relays will identify the relays which are tripped. The operator can place the relays in the proper condition and start the DGs within 10 minutes." Does the EOP l
reference a DG relay lineup procedure which is to be used by the operator? If so, what is the procedure, how does the operator access it, and how was it verified to contain the appropriate information required for these actions? How was the determination made that an operator could get to the diesel building, as'sess the i
l conditions, place the affected relays in their proper position, and start the DGs in 10 minutes? What, if any, harsh environmental conditions would an operator encounter in accessing the DGs? How was this determined?
GPC Resnonse:
Emergency operating procedures require operators to confirm automatic actuation of equipment, including the DGs. Indication in the main control room (MCR) is provided should the DGs fail to start, and abnormal operating procedure, " Diesel Generator Recovery," is implemented. There is also indication in the MCR that the relays discussed in section 2.3.1 have tripped. Both the annunciator response procedure and the DG recovery procedure direct the operator to reset the relays at the local panel in the DG building. EOPs, abnormal operating procedures, and i
annunciator response procedures are available in the MCR.
l To verify plant operating procedures contain the appropriate information for the required actions, the procedures were reviewed during development of the SSELs and again during the Operations Department review of the SSELs. The 10-minute timeframe for the operator action was determined from interviews with operators HL-5227 E-12
Response to Request for Additional Information:
U.nresolved Safety Issue (USI) A-46 and operating experience. Also, note that consideration ofloss of coolant accidents (LOCAs) and high energy line breaks (HELBs) are not required by the A-46 program as discussed in the following paragraph. For the purpose of the A-46 program, the most rapid reactor depressurization assumes automatic actuation of the automatic depressurization system (ADS) in the alternate shutdown path.
Consequently, the ac powered low pressure injection systems, and therefore the DGs, will not be needed until approximately 22 minutes into the event (Reference 1). Using either the safety relief valves for manual depressurization in the alternate path or the de powered high pressure coolant injection system (HPCI) in the primary shutdown path allows significantly more time before use of the ac powered low pressure injection systems is necessary.
Section 3.2.5 of the SQUG GlP states that the only potential events postulated to occur, other than a design basis earthquake (DBE), is a loss of offsite power (LOSP). Other events that could cause harsh environmental conditions, such as LOCAs, HELBs, and fires, do not have to be considered for the USI A-46 program.
Therefore, the only harsh environmental conditions that must be considered are those associated with a DBE and an LOSP. For Plant Hatch, the only issue identified is the possible loss of the normal lighting system, in which case the operators rely upon emergency lighting or hand-held lights to perform their duties.
- 10. NRC Request:
Section 2.6," Plant Operations Department Review of Safe Shutdown Equipment List" states in part that, "With the incorporation of the Operations Department comments concerning manual operation of valve lEl 1-F008 (2El 1-F008), it was concluded that the plant procedures and operator actions and training are adequate to direct the plant to the safe shutdown path using only equipment on the SSEL."
Section 2.6 also states, that "A crew of plant operators were observed as they performed the simulated LOSP transient to determine whether any problems were encountered." What was the shift composition of the crew which performed the simulation? Are other operating crews scheduled for or have completed similar training? How does plant management expect to train other operations crews in the expected methods to accomplish safe shutdown of the plant in the event of a DBE7 How was the item identified and dispositioned by a simulator crew?
GPC Resnonse:
The SQUG GIP states that the plant Operations Department review is intended to verify existing normal ud EOPs are compatible with the SSEL. Also, the review verifies a trained opera:or, following existing plant procedures, will eventually be directed to use the safe shutdown equipment and instruments, even though the HL-5227 E-13
r I
Response to Request for Additional Information:
l U.nresolved Safety Issue (USI) A-46 a
operator may have first tried to shut down the unit using equipment not included on the SSEL.
1 t
Section 3.7 of the SQUG GIP suggests three altemative methods to perform this review; however, plants are only required to perform the review using a single method. Plant Hatch chose to perform the review using two of the methods outlined in the GIP: 1) a " desk top" review of the applicable procedures, and 2) an LOSP scenario conducted on the simulator.
The simulator exercise served as a review of the SSEL and plant procedures as described above. The exercise was not conducted to provide operator training. All reactor operators receive simulator training as part of maintaining their licenses.
Simulator scenarios are developed to train operators on a variety of plant transients and accidents using multiple equipment failure sets. No specific scenario results from a DBE. The operating crew that performed the simulation discussed in Summary Report section 2.6 was most likely the crew undergoing simulator training at the time of the review.
As stated in the SQUG letter from Mr. Niel Smith to Mr. James Partlow of the NRC dated August 21,1992, SQUG's understanding of the NRC Staff's position on operator training is that appropriate training on plant procedures is required only when the subject plant procedures are changed to achieve compatfoility with the i
SSEL (as described in SSER No. 2, Section 11.3, Evaluation and Conclusion, Item 2). Training should be provided only to the extent necessary to familiarize i
operators with changes to the procedures as a result of the A-46 program. No additional training on existing normal shutdown procedures or symptom-based j
EOPs is considered necessary. Also, it is not necessary for operators to know the 4
items of equipment on the SSEL. Plant Hatch did not develop any new operating procedures or modify any existing operating procedures as part of the A-46 program.
The last sentence in the NRC's request questioned the identification and disposition of a specific " item." The only " item" identified by the Operations Department review was the discovery that manual action of valve 1/2El 1-F008 was not specifically addressed in either the normal or the emergency operating procedures.
j As described in Section 2.6 of the USI A-46 Summary Reports, when the scenario was performed in the simulator, the operating crew requested the valve be manually opened. As a standard practice, operators are trained to request manual action be performed if accessible equipment does not remotely operate. However, since this manual action was not specifically addressed in the procedures, the alternate shutdown path was revised to show the use of the alternate shutdown cooling (ASDC) mode of RHR rather than the shutdown cooling (SDC) mode. This revision to the alternate shutdown path eliminates the need to open the 1/2El1-F008 HL-5227 E-14
' Response to Request for Additional Information:
Unresolved Safety Issue (USI) A-46 valve. Therefore, even though the simulator exercise demonstrated that operator training was adequate to ensure the manual action was performed, this action is no longer required for use of equipment on the SSEL.
j 1
- 11. NRC Reauest:
Section 3.5, " Summary of Results" contains Table 3-1, " Summary of Relay Chatter Evaluation Results." The Table 3-1 indicates that 106 and 92 combinations of relays in Unit I and 2 were resolved by operator actions. How were these operator actions verified and validated to insure that under the postulated conditions of a DBE they could be adequately performed? What field and control room simulator scenarios were developed to verify and validate that these operator actions could be accomplished in the timeframe required to facilitate safe shutdown? How were potentially harsh environmental conditions factored into these analyses?
l GPC Resnonse:
~
The numbers given in Table 3-1 represent the relay / component combinations resolved by operator actions. The number of actual manual actions is very small by j
comparison. With the exception of the DG relays, all operator actions are performed from the control room. The rebility ofplant operators to deal with the human actions required to bring the plant to a safe shutdown condition following a DBE was given special consideration. The mechanical and electrical systems engineers worked closely with Operations Department representatives during development of the SSEL and while performing the relay chatter evaluation to ensure the number of required human actions is reasonable.
Following completion of the SSEL and the relay evaluation, the Operations Department review of the normal and emergency operating procedures was conducted to ensure compatibility with the SSEL. A review of operator actions that may be required due to relay contact chatter was also performed. The reviews confirmed that all A-46 operator actions are supported by existing nonnal or emergency operating procedures. (See GPC's response to RAI No. 9 for more details on the DG relays). All plant operators are fully trained in these procedures.
No new or abnormal modes of operation were considered and no changes to operating procedures were required. As described in the response to RAI No.10, the Operations Department review consisted of a " desk top" review of the applicable procedures and an LOSP scenario conducted on the simulator. A list of applicable operator actions is included in Appendix G of Reference 10.
Plant Hatch was the first plant to conduct a combined SMA and USI A-46 evaluation. The Hatch SMA was an EPRI pilot plant program for a BWR and a soil site. The SSEL was developed to envelope the requirements of both the SMA and HL-5227 E-15
Response to Request for Additional Information:
Unresolved Safety Issue (USI) A-46 the A-46 programs. Minor revisions to the SSEL and relay chatter evaluation results developed for the pilot plant study were incorporated based on subsequent changes in the SQUG GIP. As previously stated, the NRC performed an extensive review of the Plant Hatch programs, as documented in References 1-5. In Reference 4, the NRC Seismic Design Margins Working Group had the following comments concerning the Hatch pilot plant study:
"The Working Group acknowledges the exemplary manner in which this review was conducted...We feel the Hatch review sets a high standard for future seismic margins reviews."
As part of the NRC's review at Plant Hatch, a simulator exercise with SSEL components was arranged for NRC Staff, PRG members, and consultants. In Reference 5, which contains the results of this review, the following information is presented:
"The last part of the meeting consisted of a simulator exercise arranged by SCS/GPC. The primary purpose of the exercise was in response to concerns raised by the PRG relative to the application of procedures and 1
the viability of manual actions during a seismic event. In order to j
provide a stringent test of procedures, GPC disabled the following equipment in the simulator logic: 3 of 4 RHR pumps,1 of 2 core spray i
trains,3 of 4 RHR service water pumps,3 of 4 plant service water i
pumps, drywell spray system, automatic turbine trip system, condensate storage tank, and 7 of 11 safety / relief valves. In addition, a small break was programmed into the simulator, and manual operation of diesel
)
[
relays was assumed to be required. The simulator operating crew was informed only that a seismic event was to occur...The simulator exercise
}
demonstrated the ability of the operators to bring the event under j
l control, and transfer the plant to a safe shutdown condition."
9 l
{
Considering these operator actions under the postulated conditions of a DBE, the issue of operator response to spurious alarms was also considered. The two j
simulator exercises conducted at Hatch for the A-46 SSEL did not result in spurious alarms having an adverse effect on operator responses. The effect of spurious i
alarms is also discussed in detail in Section 3.5.3 of EPRI NP-7148-SL (Reference 11). The EPRI report summarizes this issue as follows:
I i
"Accordingly, there appear to be no reasonable bases of evidence which would suggest that spurious alarms resulting from an earthquake may j
lead to abnormal operator responses."
1
- HL-5227 E-16 f
y
..r.,
..m.
s
i j
Response to Rcquest for Additional Information:
Unresolved Safety Issue (USI) A-46 4
As stated in GPC's response to RAI No. 9, Section 3.2.5 of the SQUG GIP states a
i that the only potential event other than a DBE, postulated to occur, is an LOSP.
Other events that could cause harsh environmental conditions such as LOCAs, HELBs, and fires do not have to be considered for the USI A-46 program.
4 Therefore, the only harsh environmental conditions that must be considered are those that are associated with a DBE and an LOSP. For Plant Hatch, the only issue identified is the possible loss of the normal lighting system, in which case the
}
operators rely upon emergency lighting or hand-held lights to perform their duties.
- 12. NRC Reauest:
i j'
For the operator actions specified above, what modifications to existing operating i
procedures or development of new procedures (normal, abnonnal, and emergency) were required and what methods were used to verify and validate that these procedures are appropriate to the circumstances?
GPC Resnonse:
i i
As stated in Section 2.6 and Appendix B of the Plant Hatch USI A-46 Summary Reports, Plant Hatch procedures, operator actions, and training are adequate to direct the plant to the safe shutdown paths using only equipment on the SSEL.
l Therefore, Plant Hatch was not required to develop any new operating procedures 1
or modify any existing operating procedures as part of the A-46 program.
]
1 1
- 13. NRC Reauest:
I Provide a list of components which did not meet the wording of a caveat and were l
not treated as outliers, along with their resolution.
GPC Reso0DE:
Plant Hatch was the first plant to conduct a combined SMA and USI A-46 evaluation. The original revision of the GIP did not require specific instances to be i
documented where the wording of a caveat was not met; however, the equipment j
was not treated as an outlier because it met the intent of the caveat. The Hatch j
Unit I and a major portion of the Unit 2 combined SMA and A-46 walkdowns were performed prior to the incorporation of this requirement into the GIP. The Hatch SRTs documented on the Screening Evaluation Work Sheets (SEWS) decisions concerning an SSEL item with regard to caveats. Therefore, the requested list does not exist as a stand alone document. The list could be constructed by a thorough and lengthy review of the walkdown packages; however, this effort would involve considerable time and resources. The Hatch combined SMA and A-46 program, including the quality of the walkdown packages and documentation, has been used HL-5227 E-17
i Response to Request for Additional Information:
l U.nresolved Safety Issue (USI) A-46 1-by the industry and the NRC as a model for conducting these evaluations. The additional time and expense associated with compiling this information do not appear to be warranted. The Hatch walkdown packages are available for further 4~
review at the SNC Corporate Offices.
i l
- 14. NRC Reauest:
I.
~
i Provide a tentative schedule for completion of pending maintenance work orders (MWOs) an/or design change requests (DCRs) with regard to resolution of outliers i
identified in Appendix L of the summary report.
!~
j GPC Resnonse:
j.
As stated in Section 4.8 of the Unit 1 and Unit 2 USI A-46 Summary Reports and in the completion letter from Georgia Power Company to the NRC (Reference 12), all modifications required to resolve A-46 outliers were completed by the end of 1995.
i
REFERENCES:
p 1.
Letter from J. T. Beckham, Jr., Georgia Power Company, to USNRC dated May 30,1995, transmitting two USI A-46 Summary Reports for Edwin I. Hatch Nuclear Plant, Units 1 and 2, in response to GL 87-02.
j 2.
" Seismic Margin Assessment of Edwin I. Hatch Nuclear Plant, Unit 1,"
i EPRI NP-7217-SL, Electric Power Research Institute, Palo Alto, CA, June 1991.
l 3.
Letter from Dan Guzy (RES Co-Chairman Seismic Design Margins Working Group) to U. S. Nuclear Regulatory Commission distribution, " Resolution and l
Closure of All Soils Issues in the Hatch Review," April 29,1990.
]
4.
Letter from Dr. Michael P. Bohn (Sandia National Laboratories) to Dr. Nilesh Chokshi (U. S. Nuclear Regulatory Commission) with enclosed report,
" Independent Evaluation of the Hatch Seismic Margin Assessment - Seismic Building Response and Floor Spectra," July 5,1991.
i 5.
Memorandum from Dan Guzy (RES Co-Chairman Seismic Design Margins l
Working Group) and Goutam Bagchi (NRR Co-Chairman Seismic Design Margins l
Working Group) to U. S. Nuclear Regulatory Commission distribution, " Final Evaluation of the Hatch Seismic Margins Review," May 2,1990.
i I
HL-5227 E-18 a
j i
I
' Response to Request for Additional Information:
Unresolved Safety Issue (USI) A-46
REFERENCES:
(Continued) 6.
Letter from P. R. Davis (PRD Consulting, Chairman, Hatch Seismic Margin Assessment Peer Review Group) to Dan Guzy (RES Co-Chairman Seismic Design Margins Working Group) with enclosed report, " Hatch SMA Peer Review Group Final Report: Evaluation of the Application of the NRC and EPRI Seismic Margins Methodologies," May 3,1990.
7.
"A Methodology for Assessment of Nuclear Power Plant Seismic Margin,"
EPRI NP-6041-SL, Electric Power Research Institute, Revision 1, Palo Alto, CA.
August 1991.
j 8.
" Code Requirements for Nuclear Safety Related Concrete Structures", American Concrete Institute, ACI 349.
9.
" Recommended Approaches for Resolving Anchorage Outliers," Final Report, EPRI TR-103960, Electric Power Research Institute, Palo Alto, California, I
June 1994.
- 10. Letter from J. T. Beckham, Jr., Georgia Power Company, to USNRC dated January 26,1996, transmitting the Plant Hatch Units 1 and 2 Individual Plant Examination of External Events, Response to Generic Letter 88-20, Supplement 4.
I1. " Procedure for Evaluating Nuclear Power Plant Relay Seismic Functionality," Final Report, EPRI NP-7148-SL, Electric Power Research Institute, Palo Alto, California, December 1990.
- 12. Letter from J. T. Beckham, Jr., Georgia Power Company, to USNRC dated July 31,1996," Completion of Actions Required Per Supplement I to Generic Letter 87-02."
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