ML20072B870
| ML20072B870 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/10/1983 |
| From: | Bauer E PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8306140606 | |
| Download: ML20072B870 (38) | |
Text
{{#Wiki_filter:% W-o 4 PHILADELPHIA ELECTRIC COMPANY 2301 MARKET STREET P.O. BOX 8699 PHILADELPHI A. PA.19101 '"",^",[;"^"'""* 12isi e4i-4ooo s.. EUG ENE J. BR ADLEY assocent..wm. mat couns.6 OON ALD BLANKEN CUDOLPH A. CHILLEMI
- 3. C. MIM M H A LL.
T. H. M AHER CORN ELL '^"'^!."."".^*.."....s., June 10, 1983 ss s CDW ARD J. CULLEN, JR. THOM AS H. MILLER. J R. IK EN E A. McMENN A .ssesvant covas.L Docket Nos. 50-352 50-353 Mr. A. Schwencer, Chief Licensing Branch No. 2 Division of Licensing U.S. Nuclear Regulatory Commission
Subject:
Limerick Generating Station (LGS)-Units 1&2 Open Items from NRC Draf t Safety Evaluation Report (DSER) - Structural Engineering Branch (SEB)
References:
Letter from A. Schwencer to E. G. Bauer, Jr., dated March 11, 1983
Dear Mr. Schwencer:
Attached please find our response to SEB-DSER open item numbers 1, 4, 8, and 9. These open items were transmitted to us via the referenced letter. With these responses, all SEB-DSER open items addressed in the referenced letter have been closed-out. Response item numbers 1, 4, and 8 represent our formal response while item 9 indicates the FSAR page changes which will be made in Revision 21 (June 1983). Very truly yours, 8306140606 830610 PDR ADOCK 05000352 E PDR l Attachments Copy to: See Service List IlI
o + cc: Judge Lawrence Brenner (w/ enclosure) Judge Richard F. Cole Judge Peter A. Morris Troy B. Conner, Jr., Esq. Ann P. Hodgdon, Esq. Mr.-Frank R. Romano Mr. Robert L. Anthony Mr. Marvin I. Lewis Judith A. Dorsey, Esq. Jacqueline I. Ruttenberg Thomas Y. Au, Esq. Mr. Thomas Gerusky Director, Pennsylvania Emergency Management Agency Steven P. Hershey Charles W. Elliott, Esq. Donald S. Bronstein, Esq. Mr. Joseph H. White, III David Wersan, Esq. Robert J. Sugarman, Esq. Martha W. Bush, Esq. Atomic Safety and Licensing Appeal Board Atomic Safety and Licensing Board Panel Docket and Service Section i l i,
r ENCLOSURE 1 Response to LGS DSER-SEB Open Item Numbers 1, 4, 8, and 9 SEB (220) DSER #(1) Response Spectra Frequencies (3.7.1) In the design of plant structures, systems and components, an operating basis earthquake (OBE) of.075g horizontal and a safe - shutdown earthquake (SSE) of 0.15g horizontal were used. The F values for the vertical component of the design response spectra are 2/3 of the horizontal values described above. The response spectra are based on data developed from records of previous earthquake activity and represent an envelope of motion expected at a sound rock site from a nearby earthquake. Comparison between the LGS design response spactra and R.G. 1.60, " Design Response Spectra for Seismic' Design of Nuclear Power Plants", indicated that for frequencies between '. he and 5 ha, R.G. 1.60 exceeds the LGS design spectra. The applicant should discuss the significance of these exceedances on structures, piping, equipment and systems essential for the safe shutdown of'the-plant. The staff considers this to be a confirmatory item. Response : The LGS design _ response spectra (Figure' 3.7-1), in conjunction with the damping values shown in Tables.3.7-1 and 3.7-2, forms the design basis for-LGS seismic analysis. For the seismic analyses of Category 1 structures (e.g. containment structures, reactor & control enclosures, diesel generator building, and the spray pond pump house) which contain safety related piping, equipment, and systems, damping values equal to 2% and 5% of critical are used for OBE and SSE, respectively. In accordance with RG 1.61, damping values equal to 4% and 7% of critical are acceptable for these reinforced concrete structures. For comparison, the horizontal LGS design spectra are plotted with ( the RG 1.60 horizontal spectra in ' figures SER-1-1 and SER-1-2. These plots show that both the OBE and SSE horizontal LGS design spectra envelop the RG 1.60 spectra for frequencies greater than 1.0 hertz. The vertical LGS design spectra are equal to two-thirds of the horizontal, per FSAR section 3.7.1.1. For comparison, the vertical i LGS design spectra are plotted with the RG 1.60 vertical spectra on figures SER-1-3 and SER-1-4. These plots show that the RG 1.60 vertical spectra are generally higher than the LGS design spectra. i t i
r. In the WASH-1255 report (Reference 1) Newmark has shown, based upon fourteen strong seismic motion records,.that the ratio of vertical to the-horizontal ground acceleration is two-thirds on the average. Although only 3 of 14 earthquakes considered were on rock, the ratio of the vertical to the horizontal acceleration is less in rock than in alluvium. Later, in the'NUREG/CR-0098 report (Reference
- 2) Newmark recommended that the vertical design motion be taken as two-thirds of the horizontal across the entire frequency range.
In addition, from the research based on the analysis of thirty vertical recordings made on "hard" or rock sites, the authors Rizzo, Shaw, and Snyder report (Reference 3) that the ratio of two-thirds between the vertical and horizontal accelerations is conservative. Their study included sites located in the Eastern United States, such as Blue Mountain Lake and New York. It is stated in their report that the RG 1.60 spectra envelop both rock and soil sites, and it is shown that the RG 1.60 provisions are overly conservative for "hard" or rock sites. Since all principal Category 1 structures (coutainment structures, reactor & control enclosures, diesel generator building, and the spray pond pump house) of the project are founded on competent rock, we have concluded, based on the above discussion, that the difference between the LGS design vertical spectra and the RG 1.60 vertical spectra has'no impact on safety-related structures, piping, equipment, and systems. The LGS design spectra, used in conjunction with the more conservative damping values, assures safe operation of the plant during a seismic event. TGS/dme 15/1 L l 1 l s.
References for Enclosure 1 1) "A Study of Vertical and Horizontal Earthquake Spectra", USAEC Contract No. AT (49-5)-2667, WASH-1255. N. M. Newmark Consulting Engineering Services, Urbana, Illinois, April 1973. 2) " Development of Criteria for Seismic Review of Selected Nuclear Power Plants", USNRC Contract No. AT (49-24)-0116, NUREG/CR-G098, N. M. Newark Consulting Engineering Services, Urbana, Illinois, May 1978. 3) " Vertical Seismic Response Spectra", P. C. Rizzo, D. E. Shaw, and M. D. Snyder, Journal of the Power Division, ASCE, January 1976.
s e ' ^* . N PER MSEC. SS 0 1.0 9.1 0 01 siae e a e i a 3i ia a i e i eis, ie i a s LEGEND TIME HISTORY RESPONSE SPECTRA ........ DESIGN RESPONSE SPECTRA I i s lll a I 5 W i 5 E O E k \\ [ G7c'ht643 I m Y V.70 P% \\ l[*5Zl'. l %Q $7 / 0.0151 sS N-h. s"f j
===;a. 0.1 2 4 6 8 1.0 2 4 S S 10.0 2 4 5 0 100 l FREQUENCY CPS \\- 3%A 4 l b C 7t fi@ k - @ E LIMERICK GENER ATING ST ATION UNITS 1 AND 2 l FINAL SAFETY ANALYSIS REPORT ( COMPARISON OF TIME HISTORY RESPONSE SPECTRA AND DESIGN l i RESPONSE SPECTRA (2% DAMPlNG) N-FIGURE
. t 2 a.. _. _ i kt PER100 SEC. 1. 1 .1 .1 Iss a a i e i a ssssa s s a a sais a i ie e a LEGEND TIME HISTORY RESPONSE SPECTR A ......... DESIGN RESPONSE SPECTRA I l Ei .s E D kh IN ~ C3% h fe4) l 'l l t 7 / / S Mb /.-7 ne ~ 1 &f / N w j 0.b g. W t t. .. io.. FREQUENCY. CPS M \\.bD MlM ( NNN k%%ML-SE LIMERICK GENERATING STATION i UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT C COMPARISON OF TIME HISTORY ( RESPONSE SPECTRA AND DESIGN l RESPONSE SPECTRA (5% DAMPING) -k-FIGURE l
PERIOO-SEC. to o 1o e.1 c.of siii e i i i i sii i i ie i isi ii ie i i LEGEND TIME HISTORY RESPONSE SPECTH&. ....... DESIGN RESPONSE SPECTRA I i n E % 1.60 MQS.m i aimwwen E // e 3 / l Ps 1 / ~ g iaL a V 70V V V/....7.. ..s.. j s ( A pr Q . -o.ow$. 'c.05 m......;;;;# t 0.1 2 4 8 8 1.0 2 4 8 8 10.0 2 4 8 e 100 FREQUE NCY. CPS kN \\hda_- N LIMERICK GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSIS REPORT COMPARISON OF TIME HISTORY RESPONSE SPECTRA AND DESIGN RESPONSE SPECTRA (2% DAMPlNG) N-k-FIGURE
d PER900-SEC. j O*## 10 0 1.0 0.1 555 5 5 5 5 I l gI g l 5 I I i 5 5 5 B 5 I LEGEND TIME HISTORY RESPONSE SPECTRA ....... DESIGN RESPONSE SPECTRA I 6 m= = E s 5 E g.M Wo S?EGte. i / CUu%nRQ g, .L f % s d - ?+w ) g;;*V / '%Q -o.\\5 g. sf '-o.\\o g. 0.1 2 4 8 8 1.0 2 4 5 0 10.0 2 4 8 8 100 PREOUENCY. CPS LGS DEGGA SeeQRe.(5. }.MW4 LIMERICK GENERATING STATION UNITS 1 AND 2 FINAL SAFETY ANALYSl3 REPORT COMPARISON OF TIME HISTORY RESPONSE SPECTRA AND DESIGN RESPONSE SPECTRA (5% DAMPINU) EQ_-(- k) FIGURE
DSER #4 - Containment Dynamic Modal (3.7.2) In additicn to the problems with the rock parameters mentioned in (3) above, a structural design audit performed by the staff revealed problems with the containment dynamic model. These problems consist of errors in calculating the containment diaphragm slab spring stiffness and mass distribution of the containment building floors. Tha staff considers these problems to be open issues.
Response
(A) Spring Stiffness of Diaphragm Slab The diaphragm slab is rigidly connected to the containment wall and the reactor pressure vessel (RPV) pedestal wall. Hence, the equivalent spring representing the diaphragm slab in the vertical containment model is calculated using a fixed-fixed boundary condition. The error
- in calculating the equivalent spring stiffness has been corrected.
The containment vertical uncracked seismic model has been reanalyzed with the new corrected spring stiffness for the SSE event. Comparisons of modal frequencies, modal participation factors, and selected acceleration response spectra between this revised spring model and the design model are presented in attachment 1. From these compariscus, it is concluded that: a. Adjustment of the spring stiffness of the diaphragm slab induced negligible variation in the structural response for the containment structure, and b. The current seismic analysis of the containment structure is adequate. (B) Mass Distribution of Platforms within the Drywell of the Containment Structure All platforms spanning between the containment wall and the RPV pedestal wall in the drywell are supported in the horizontal direction only at one end and are free to slide at the other end. They are supported at both ends in the vertical direction. Current horizontal and vertical seismic models of the containment structure assumed a platform mass distribution of 80% to the containment wall and 20% to the pedestal wall. A study has been performed using the uncracked containment models to determine the effects of different distribution of this platform mass. The horizontal study model considers 100% of the platform masses lumped to the restrained ends and zero mass to the unrestrained ends. The vertical study model redistributes the platform masses to the containment wall stick and to the RPV pedestal stick of the model according to tributary area (approximately 60% to the containment wall and 40% to the pedestal wall).
- Discovered during the NRC-SEB audit of Bechtel in October of 1982.
The study models were analyzed for the OBE event and the results_ compared to the design values. These comparisons are shown in attachment 2 for the horizontal model and in attachment 3 for the vertical model. These comparisons show that the distribution of platform masses have negligible effect on the structural response of the containment structure. Therefore, the current seismic analysis of the containment structure is adequate. TGS/dmc 15/2 i i l l l
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- DSER #8 - Reactor / Control Building Load Path Assumptions (3.8.3)
Seismic Category 1 structures other than the containment and its interior structures include the reactor / control building, diesel generator building, spray pond pump structure and the spray pond. All of the aforementioned structures are of concrete and structural steel. During a structural design audit of the reactor / control building calculations by the staff, it was learned that the horizontal load transfer assumptions between the roof / floor diaphragms, the shear walls and the foundation were not cicar. To verify the logic of the load path assumptions relied upon, the applicant should provide a description of the assumptions made and verification of agreement between the calculations and the assumptions. The staff considers this to be a confirmatory item.
Response
In the design of shear walls, the following assumptions are made: 1) All exterior and interior bearing reinforced concrete walls (parallel to the direction of the earthquake) are considered to be subjected to earthquake forces and are accounted for in the stress analysis. Perpendicular walls are included in the analysis for horizontal torsion. 1 2) A floor is considered to be a rigid diaphragm if the entire floor has a reinforced concrete slab and the total horizontal forces at any level are assumed to be distributed to the vertical resisting elements (shear walls) in proportion to their relative rigidities. 3) Where a shear wall has openings, each pier between the openings is assumed to be tied at the top and the bottom of the openings by a stif f spandrel beam or by a foundation at the bottom. The point of inflection is assumed to be at the mid height of the pier. The total lateral force on the wall is distributed to the piers in proportion to their relative rigidities. 4) If the openings are close together in a row, the piers between I them are assumed to be incapable of transfering lateral seismic forces. 5) Where the floor slab is interrupted by large openings, such as I the containment structures inside the Reactor Enclosure, a partial diaphragm configuration is considered. Only the shear walls connected to the partial diaphragm are considered for the lateral seismic force distribution. Crating floors are not considered to be acting as a diaphragm.
6) At each main floor level, each shear resisting element in a shear wall is checked for the assigned forces for shear and bending. For the Reactor / Control Enclosures the entire structure is considered as an integral multi-celled box structure and is checked for overall overturning moment due to lateral seismic forces in each (N-S or E-W) direction combined with the vertical seismic forces. The calculated stresses (tension and compression) from this procedure are then combined with stresses due to local bending in each of the shear resisting elements, including the piers, for the checking of concrete bending and shear stresses and designing of reinforcing steel requirements in accordance with the allowable values in ACI-318-71 code. 7) An illustrative example of the above assumptions and prccedures is presented in the attachment. Walls on Col. line 19.4 and 26.6 between elev. 200/201' and 217' at the Reactor / Control Enclosures have been analyzed for the N-S earthquake. TGS/dmc 15/3
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DSER #9 - Spent Fuel Pool (3.8.3) In the original spent fuel pool (SFP) design, the applicant assumed low density storage racks would be used and the SFP floor was designed accordingly. Revision 10, of the Limerick FSAR mentions the use of high density free standing storage racks. The applicant should provide verification that the SFP floor retains sufficient structural capacity to withstand all loads, load cases, and load combinations relied upon in the original calculations. The staff considers this to be a confirmatory item.
Response
The LCS FSAR will be changed in the June 1983 revision to read as shown in Attachment A. I 4 I r
ATTACHMENT "A" Sheat l'of 2 LGS FSAR energy. These impact loads have been verified by full-size tests on an actual top grid casting. For condition 3, an unimpeded fuel assembly drop through an empty cavity, an equivalent static load was determined to shear out the bottom fuel support. The following presents the equivalent static loads for the three drop conditions. Condition Description Load 1 36 inch drop, corner of rack 82,500 lb 2 36 inch drop, middle of rack 64,300 lb 3 Drop through empty cavity of 24,600 lb rack Conditions 1 and 2 are the loads due to vertical impact. The subsequent roll over impact load was shown to be less than the above stated vertical impact values. Equivalent static loads for different dropped fuel bundle cases were applied at proper locations to the ANSYS finite element i model of the rack and combined with the dead weight vertical load (rack l full of fuel). Stresses for each member and plate were tabulated and i were less than the factored allowables of Equation 4, Table 9.1-20. I l 9.1.2.3.2.4 Pool Interface Loads The spent fuel rack load acting on the spent fuel pool is determined j using the results from the ANSYS seismic model described in Section 9.1.2.3.2.2.a and shown in Figure 9.1-36. The force response at the i rack and pool interface elements of the two-rack model are computed for each time step of the dynamic analysis. These force responses which include SSE loading are used to determine the maximum concentrated local i forces and the equivalent uniform global rack load acting on the spent fuel pool slab. The maximum pool slab bearing and punching shear stresses are computed from the maximum spring forces at nodes 101 and 102 including the impact effect. The equivalent uniform rack load is conservatively computed using the maximum reaction force acting at either nodes 201, 202, 203, or 204. 9.1-26. I ,.,.-p..,2 --Fv+ w-+ r--'-w.--.*-~m+w,*--e*- w----=
- -w=*1 ea
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ATTACHMENT "A" Sleet 2 of 2 = LGS FSAR The : allowable stresses for bearing and punching shear of the spent fuel pool slab are in accordance with sections 10.14 and 11.10, respectively, of the building code requirements for reinforced concrete (ACI 318-71).- the allowable uniform rack load on the pool slab is calculated to be 2815 psf, using a finite element model at the spent fuel pool. The controlling load combinations are normal operating and severe environmental as given in Table 3.8-9. The assessment results given below include the additional conservatism of substituting SSE values for the OBE loads. Fuel pool slab Allowable interface load (including SSE) Condition Bearing on Concrete-Floor 4760 psi 4748 psi (Local Area) Punching Shear-Floor 253 psi 132 psi Uniform Load-Floor Slab 2815 psi 2811 psf (Seismic) 9.1.2.3.2.5 Summary and Conclusions A 5.50-inch minimum clearance is maintained between the rack and any pool wall or obstruction to avoid rack impact during a seismic event. All member and plate stresses satisfy the stress combination limits and factored allowable stresses of Equations 1 through 6, Table 9.1-20, for the seismic and dropped fuel conditions. The stresses on the concrete floor for both seismic and dropped fuel conditions are acceptable in accordance with the allowable limits. 9.1.2.3.3 Installation of New High Density Racks Racks are lifted individually from the refueling floor laydown area and lowered into position in the pools using the 125-ton reactor enclosure crane and a remotely actuated pneumatic lifting device provided by the rack vendor. Each rack is aligned and leveled as it is placed in its proper-pool. Each rack is a freestanding unit that rests on the pool floor using four bearing pads attached to corner leveling screws. 9.1-27 1TGS/dmc 15/4}}