View source for 3F0997-01, Submits Suppl to NRC Ltr Re Protection Against Dynamic Effects of Loca.In Addition,Issue Being Tracked by NRC as Inspector Followup Item 50-302/95-15-02.W/one Oversize Drawing

{{Adams
| number = ML20217C278
| issue date = 09/29/1997
| title = Submits Suppl to NRC Ltr Re Protection Against Dynamic Effects of Loca.In Addition,Issue Being Tracked by NRC as Inspector Followup Item 50-302/95-15-02.W/one Oversize Drawing
| author name = Rencheck M
| author affiliation = FLORIDA POWER CORP.
| addressee name = 
| addressee affiliation = NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
| docket = 05000302
| license number = 
| contact person = 
| document report number = 3F0997-01, 3F997-1, 50-302-95-15, TAC-M96604, NUDOCS 9710010330
| title reference date = 06-05-1997
| document type = CORRESPONDENCE-LETTERS, INCOMING CORRESPONDENCE
| page count = 16
| project = TAC:M96604
| stage = Other
}}

=Text=
{{#Wiki_filter:. . . .    . . . . .
we Florida fgggpr
    *:.0: Je September 29, 1997 3F0997-01 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D. C. 20555-0001

==Subject:==
Protection Against Dynamic Effects                       of      LOCA,       IFI     95-15-02, Supplemental Letter (TAC No. M96604)

==References:==
: 1. FPC to NRC letter, 3F0697-13, dated June 5, 1997 l                             9. NRC to FPC letter, 3N0497-12, dated April 10, 1997
)

==Dear Sir:==

l           Florida Power Corporation (FPC) is hereby submitting this letter as a supplement to Reference 1. In that letter, FPC committed to provide additional information to address the effects of jet impingement between the Crystal River Unit 3 (CR-3)
Pressurizer Surge Line and the Nuclear Services Closed Cycle Cooling (SW) System lines serving the Reactor Coolant Pumps (RCP) by September 29, 1997. In addition, this issue is being tracked by the NRC as Inspector Followup Item (IFI) 50-302/95-15-02.
As FPC stated in Reference 1, two items remained open for FPC to complete so IFI                                       '
95-15-02 could be resolved.             FPC indicated that further evaluation of jet impingements was necessary to confirm whether these possible interactions created adverse effects between the surge line and the SW System piping. In paragraphs d and e of Reference 1, FPC stated the following:
: d.         Circumferential pipe ruptures do have the potential for jet impingement on SW piping associated with RCP-1A because of the free stream expansion jet produced by a circumferential pipe rupture and there are no jet
                                                                                                                               /f
                                                                                                                                ,/
impingement barriers between the surge line and the SW                                                 >
piping.
9710010330 970929 PDR    ADOCK 05000302                                             lg((((((((,(({\{}{hkkkkk I, ,,.c           *-
l W']/
F                             PDR CRYSTAL RIVER ENEROY COMPLEX: 15780 W. Power une St e crystar mv.,. Fiorus 3442s e70s e 1362: 7es-s4ae A hande Progress Conneeny
                                              )V4 L   @              h        8-    I    {

       . . U. S. Nuclear Regulatory Commission 3F0997 01
  .         Page 2 of 2
                -    e. Longitudinal pipe ruptures also have the potential to result in jet impingement on SW piping associated with RCP-1A and RCP-1B because of the free stream expansion jet produced by a longitudinal pipe rupture and no jet impingement barriers between the surge line and the SW piping.
FPC has completed that evaluation of the potential impact of the jet impingement due to a circumferential or a longitudinal pipe rupture. FPC has concluded that neither a circumferential nor a longitudinal pipe rupture will adversely affect the SW System piping serving the RCPs. The results of the evaluation are presented in Attachment A.
Based on this evaluation, FPC determined that Final Safety Analysis Report (FSAR) changes will be required to clarify the licensing basis for ruptures on the pressurizer surge line. First, FSAR Section 4.2.6.6 will be revised to state that longitudinal ruptures were eliminated by following the guidelines of Generic Letter 87-11 " Relaxation in Arbitrary Intermediate Pipe Rupture Requirements."
Second, FSAR Section 14.2.2.5.11 will be revised to state that the fluid jet created by a circumferential rupture is limited to a 20' included angle consistent with Standard Review Plan 3.6.2, " Determination of Rupture Locations and Dynamic Effects Associated With the Postulated Rupture of Piping," and ANSI /ANS 58.2-1980, " Design Basis Protection of Light Water Nuclear Power l            Pl ant s . "
l FPC has completed a Safety Assessment /Unreviewed Safety Question Determination (SA/USQD) on these changes to the CR-3 licensing basis. There are no Unreviewed Safety Questions as defined by 10 CFR 50.59. A copy of the proposed FSAR changes is provided, for information only, in Attachment B. The final form of the FSAR change will be submitted to the NRC in FSAR Revision 24 on or before December 8,
             .997.
There are no adverse interactions between SW System piping serving the RCPs and the Reactor Coolant System Piping including the pressurizer surge line. The results presented in this letter and Reference 1 will allow FPC to close Restart Issue D-58, " Design Requirements for Dynamic LOCA Effects." The D-58 closure package will contain copies of the FPC letters discussing the issue, internal FPC correspondence related to this issue, and those FPC calculations used to resolve the issue. This material is available for NRC inspection.
The commitments in this letter are contained in Attachment C.
If you have any further questions, please contact Mr. David Kunsemiller, Manager, Nuclear Licensing at 352-563-4566.
Sincerel jfWhy.eudnL M. W. Rencheck, Director Nuclear Engineering and Projects MWR/jwt Attachments xc:      Regional Administrator, Region II Senior Resident Inspector NRR Project Manager

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 3F0997-01 ATTACHMENT A EVALUATION OF JET IMPINGEMENT EFFECTS ON CR-3 PRESSURI7sER SURGE LINE i
                                                                        - _m___-

                                                                . =  .
                                                                                      =                         .,
U. S. Nuclear Regulatory Commission                                    Attachment A 3F0997-01                                                               Page 1 of 8 Introduction In Reference 2, noted in the cover letter, NRC requested FPC to document the results and resolution of any dynamic loss-of-coolant-accident (LOCA) effects on the Nuclear Services Closed Cycle Cooling (SW) System. Reference 1 provided the results and resolution for the dynamic LOCA effects due to pipe whip. Dynamic i         LOCA effects due to jet impingements were the only remaining issues to be resolved. This attachment describes the results of FPC's evaluation of jet impingement effects on the SW System lines serving the Reactor Coolant Pumps l         (RCP).
Current licensina Basis The licensing bases for potential impact of the jet impingement due to a circumferential or a longitudinal pipe rupture is defined in Final Safety Analysis Report (FSAR) Section 4.2.6.6, "LOCA Restraints," and FSAR Section 14.2.2.5.11, " Reactor Building Subcompartments Pressure Response."       The bases are:
a       Postulated circumferential and longitudinal pipe ruptures are possible anywhere within the Reactor Coolant System pressure boundary, a       The break is assumed circular, with an area equal to the cross-sectional area of the pipe, a       The fluid jet is assumed to be conical with an included angle of 30'.
I e       Initial thrust is given by the initial pipe pressure multiplied by the cross-sectional area of the pipe.         The pressure at any impingement plane is assumed to be indirectly proportional to the                         j i
cross-sectional area of the conical expansion being generated,                            i s
l'        These licensing bases apply to the hot leg and cold leg Reactor Coolant System (RCS) piping. These are large bore pipes which involve the Leak-Before-Break (LBB) criteria and General Design Criteria (GDC) 4 exemption previously approved by the NRC for use at CR-3. As the NRC indicated in Reference 2, the RCS branch lines, such as the pressurizer surge line, do not fall within the scope of the LBB criteria and GDC-4 exemption. The present RCS piping design documentation for the CR-3 pressurizer surge line is not of sufficient detail to judge the extent of the application of two of the current licensing bases to the surge line dynamic interactions -- longitudinal breaks and jet fluid angle.              FPC has reviewed the NRC and_ industry guidance on the application of methods to ensure that jet impingements from high energy lines do not interact adversely. NRC guidance which was developed after most of the CR-3 RCS piping design and analyses were performed do provide more definitive criteria for considering dynamic LOCA' effects.
                                                            ~-. .

U. S. Nuclear Regulatory Commission                                          Attachment A
* 3F0997-01                                                                      Page 2 of 8 The use of these later NRC guidelines will require FSAR revisions. These revisions are- necessary- to clarify the licensing bases for ruptures of the pressurizer surge line to be consistent with current NRC guidelines and industry analytical methods. Thai bases being revised deal with postulated longitudinal pipe rupture and fluid jet angle. The basis for each of these changes is
,           described in the following paragraphs. As pa.t of this research, the CR-3 Operating License Safety Evaluation Report (SER) and its four supplementt (References 1 through 5 located at the end of this attachment) were reviewed to confirm that the NRC had not imposed additional requirements beyond those described in the FSAR. Revisions to these bases do not involve an unreviewed safety question because these changes implement analytical techniques that are consistent with NRC guidelines and do not conflict with methodologies already reviewed and approved by the NRC for CR-3.
Lonoitudinal Pine Ruotures A postulated longitudinal pipe rupture has the potential to result in jet impingement on SW piping associated with RCP-1A and RCP-18.              This impingement could occur because of the free stream expansion jet produced by a rupture.
There are no jet impingement barriers between the surge line and the SW piping.
The guidance in Generic Letter 87-11. " Relaxation in Arbitrary Intermediate Pi)e Rupture Requirements," would allow longitudinal breaks on the surge line to )e eliminated if specific criteria were satisfied. Generic Letter 87-11, Section B.3.b(1) states
* Longitudinal breaks in fluid systems and branch runs should be postulated in nominal pipe sizes 4-inch and larger, except where the maximum stress range exceeds the 1imits speciffed in 8.1.c.(1) and B.1.c.(2) but the axial stress range is at least 1.5 times the circumferential stress range.* If CR-3 surge line stresses could be shown to meet this criteria, longitudinal breaks could be eliminated from consideration.
FPC performed analyses on the surge line to determine if stresses were in excess of the Generic Letter 87-11 limitations. To preserve conservatisms in evaluating possible surge line break locations, FPC used the CR-3 surge line stress analyses that were generated in response to-implementing the requirements of NRC Bulletin 88-11, " Pressurizer Surge Line Thermal Stratification." .The pipe rupture stress analyses show that all postulated breaks in the straight pipe sections and the elbows of the surge line are circumferential breaks. Possible longitudinal break locations on the CR-3 pressuriter surge line do not meet the Generic Letter 87_-
11, Section B.3.L(1) guidance for consideration as breaks.
Therefore, FPC will-revise the licensing basis for pipe ruptures in FSAR Section 4.2.6.6,                  "LOCA Restraints," to state that FPC analyses have eliminated longitudinal ruptures from consideration for the pressurizer surge line based on the application of Generic Letter 87-11. As stated in Generic Letter 87-11, licensees are permitted to implement the requirements of this generic letter unless such changes conflict with the license or technical specifications. FPC's review of the CR-3 Operating License and the Improved Technical Specifications indicates that they do not contain any requirements that conflict with these licensing basis changes. The proposed FSAR change to Section 4.2.6.6 is included in Attachment B.
(

U. S. Nuclear. Regulatory Commission                                 Attachment A 3F0997-01                                                              Page 3 of 8 fluid Jet Anale A postulated circumferential pipe rupture does have the potential for jet impingement on SW piping essociated with RCP-1A. This impingement could occur because of the free stream expansion jet produced by a rupture. There are no jet impingement barriers between the surge line and the SW piping. This conclusion of a potential for jet impingement on SW piping was based on the requirement for a conical fluid jet with an included angle of 30'. Figure 1 in this attachment is a simple isometric drawing that shows the surge line circumferential break locations of interest for this discussion.        Break location #4 is the most limiting break for impact on RCP-1A. This break is at the maximum distance from the target pipes where the expansion of the jet cone is the greatest. Figurc 2 (FPC Drawing 112826D) in this attachment is an isometric representation of the RCS showing the relative location of components.
FSAR Section 14.2.2.5.11 (page 14-69) discusses the jet impingement forces for large pipe break sizes (8.0 ft" and 14.1 ft*). It states "The forces on targets in the path of escaping fluids are dependent upon the size and shape of the l
target and its distance from the rupture area. Since the actual shape of the rupture dictates the flow field shape eing generated, it is assumed that a typical rupture is circular and the free stream expansion of the jet conical with an included angle of 30"." Cutoff distances (i.e., the distance from the break plane beyond which impingement effects naed not-be considered) and target shape       ,
factors would have been used to determine the population of impingement targets      i and effective applied impingement pressure. However, without specific details       '
from the original CR-3 design documentation to define these parameters, FPC can not document the impact of a potential break location's fluid jet effects. FPC has chosen to upgrade the CR-3 analytical methods to follow NRC and industry accepted methodologies for establishing the jet angle. These methodologies describe the criteria that must be considered for evaluating fluid jets.
Much of the development of the methodolgy for analysis of dynamic effects of HELBs was a result of the Atomic Energy Commission (AEC) letter written by Mr.
A. Giambusso on December 22, 1972, that contained design guidance for HELBs outside containment and the May 1973 version of Regulatory Guide (RG) 1.46,
               " Protection Against Pipe Whip Inside Containment." The techniques used to develop jet geometry prior to issuance of Giambusso's letter were based en experiments that did not consider the jet as a steam / water mixture (fluid jet geometry for impingement is not defined in either Giambusso's letter or R.G.
              -1.46). 1he resulting jet shape was conservatively based on subcooled water or steam conditions (rather than a mixture of water / steam) and was conical with an included angle of 30', This is consistent with the jet geometry for RCS breaks referenced in FSAR Section 14.2.2.5.11.
Refinement of the CR-3 impingement jet geometry from 30' to 20' reflects developments in industry methods since the original CR-3 design.      The standard jet geometry used for the evaluation of impingement e"ects at nuclear power plants, as defined in the Standard Review Plan 3.6.2, " Determination of Rupture Locations and Dynamic Effects Associated With the Postulated Rupture of Piping,"
               ~                                                                                 -

e U. S. Nuclear Regulatory Commission                                         Attc:hment A
              . 3F0997-01                                                                    Page 4 of 8         {
and ANSI /ANS 58.2-1980, " Design Basis Protection of Light Water Nuclear Power Plants," is conical with an included angle of 20' (both Revision 0 in November 1975 based (NUREG-75/087) and Revision 1 of SRP 3.6.2 in July 1981). This shape is on analytical and empirical data and recognizes that the jet is a steam / water mixture.              In addition, for piping breaks outside the Reactor Building, CR-3 follows the methodology in FPC's " Pipe Rupture Analysis Criteric Outside the Reactor Building," that has been reviewed and approved by the NRC (NRC to FPC letter, 3N0490-10, dated April 11,1990). It also uses a 20' cone angle. The development of NRC and industry guidance since the early 1970s for fluid jet angles leads to the conclusion that the difference between the jet angles in the FSAR and those recommended today for inside and outside containment analyses results from the vintage of the CR-3 RCS piping jet impingement analyses.
The current NRC guidance in SRP 3.6.2, the CR 3 HELB Criteria, and ANSI /ANS 58.2-980 can serve as a foundation upon which t jet shape half angle of 10' (20' total) can become the basis for a CR-3 pressurizer surge line rupture. FPC has analyzed circumferential breaks on the surge line and applied a jet shape half l                 angle of 10' according to the current NRC guidance and FPC's HELB Criteria. The l                 half angle methodology is defined by ANSI /ANS 58.2-1980.
Figure 3 in this attachment is based on ANSI /ANS 58.2-1980 and should aid in the following discussion. The methodology assumes that for a circumferential break with full separation of the process piping, the jet shape is a linear increase in radius from the initial break plane to the asymptotic plane at a distance of
,                 L,. The jet area remains constant at the asymptotic plane until intersecting with the 10' half angle. From this point on, the jet area continues expanding along the path of the 10' expansion half angle.
For CR-3, the asymptotic plane, L., is no more than 88 inches from the postulated break location #4 on Figure 1. The minimum distance from the break location to the SW System lines on RCP-1A that are of interest is 159 inches. The jet shape expanding along the path of the 10' expansion half angle will miss the SW System lines by at least 8 inches for the maximum pressure conditions defined in FSAR Section 14.2.2.5.11.
Therefore, a circumferential rupture in the CR-3 pressurizer surge line will not impact the SW System lines serving RCP-1A. FPC will revise FSAR 14.2.2.5.11,
                  " Reactor Building Subcompartments Pressure Response," to state that for the pressurizer surge line the jet is conical with an included angle of 20'.             The proposed FSAR change to Section 14.2.2.5.11 is included in Attachment B.
Conclusions Longitudinal pipe ruptures on the CR-3 pressurizer surge line can be eliminated from consideration as part of the licensing and design bases in accordance with the guidance in SRP 3.6.2 and Generic Letter 87-11, therefore, jet impingements from this type of rupture will not occur. A license amendment is not required because this change does not conflict with the CR-3 Operating License or the Improved Technical Specifications.

U. S. Nuclear Regulatory Commission 3f0997-01                                                             Attachment A Page !: of 8 Circumferential pipe ruptures on the CR-3 pressurizer surge line will not produce fluid jets that impact any SW System piping serving the RCPs. The fluid jet released by the pipe rupture does not impinge on RCP-1A using the assumptions from SRP 3.6.2 and ANSI /ANS 50.2-1980.
FPC has completed a Safety Assessment /Unreviewed Safety Question Determination (SA/USQD) in accordance with FPC's 10 CFR S0.59 program on these changes to the CR-3 licensing basis in the FSAR.        There are no Unreviewed Safety Questions created by these changes.
References
: 1. CR-; Operating License Safety Eva'.Lation Report (SER), July 5,       1974, l                                3N0774 01
: 2. CR-3 Operating License Safety Evaluation Report (SER), Supplement #1, 1                                January 17, 1975, 3N0175-01
: 3. CR-3 Operating License Sr.fety Evaluation Report (SER), Supplement #2, December 3, 1976, 3N1276-09
: 4. CR-3 Operating License Safety Evaluation Report (SER), Supplement #3, December 30, 1976, 3N1276-10 j                         5.
CR-3 Operating License Safety Evaluation Report (SER), Supplement #4, January 27, 1977, 3N0177-09 o
                                                              . gr

                .         U. S.' Nucl ar Regulatory Commission                       Atthchm.nt A 3F0997                                                  Page 6 of 8 PREESURIZER NORTH CONNECTION
                                                    \'                  m a J                   ,/s,, m (SR-4 HOT tra CONNECTION
                                                       =
l i
151 i
f FIGURE 1 PRESSURIZER SURGE LINE ISOMETRIC WITH
                         . BREAK LOCATIONS

DOCUMENT PAGE(S) PULLED SEE APERTURE CARD FILES
{
(*****'***********7****f********
    * * ***   * ***  ***********f************************,*******
ACCESSION NUMBERS OF OVERSIZE PAGES:
qwas/9330-0/
                                               ,M 6    W.

U. S. Nucl:ar Regul: tory Commission                    Attachment A 3F0997-01                                                   Page 8 of 8 M
Break plane       Asymptotic plane
:  La e-
[
                                                                     'io~deg i
                                          ._g._._._.. p_._._._._._t _._._._._._._._._._._._._ _._._ ._._._. ce nie rlin e of pip e N             ____
I                                         FIGURE 3 GEOMETRY FOR A FULL SEPARATION CIRCUMFERENTIAL BREAK Source: ANSI /ANS 58.2-1980

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 3F0997-01 ATTACHMENT B PROPOSED FSAR REVISIONS i
Strikeout / Shadow Changes l
l Text being deleted         4edicated by strikccat Text being added          Indicif6dtby'ihidowing i
i    .
I

The cases of maximum impinging velocity and constant force for side split and guillotine ruptures are investigated to determine the capability of the supporting structure to absorb the energy imparted to it. The allowable limit strain in the beams is incipient strain hardening of the compression-tension, stress-strain curve, this being the general limit of unsuported beam flanges as shown in Figure 14 of Transactions, ASCE Vol. 124, 1959. This results in an allowable ductility factor of 12 for ASTM A36 steel.                    Therefore, a design utilizing plastic mechanisms to resist the rupture thrust force will give satisfactory results for the structural system.
When a shear mode of response governs the situation, Hall and Newmark nave shown that the ductility factor for compact rugged structural shapes by far exceeds 12, 4.2.6.5 Pumo and Motor The RCP casing, internals, and motor weight are supported by the 28 inch coolant lines and constant load hangers attached to the motor. In the cold condition, the RCS piping will support the RCP and motor without the hangers.                          The structural portions of the constant load hangers for the RCPs are constructed in accordance with USAS B31.7 Nuclear Piping Code and/or MSS-SP-58, as applicable.
4.2.6.6 LOCA Restraints Each steam generator has restraints located at the elevation of the upper tube
:,heet which transfer forces from the generator into the shield walls in the event of a circumferential rupture of the 36 inch line. Each 36 inch reactor coolant outlet line has a restraint located outside of and bolted to the primary shield to limit pipe motion in the event of a circumferential rupture of the piping inside the primary shield.
The design criteria which were used to examine the effects of pipe rupture did consider that postulated pipe breaks could occur at any location within the reactor coolant pressure boundary. Both longitudinal and circumferential type ruptures were evaluated. F6 rt hir iinilfs i siTd ocuins6tsdt i n* F PC TC al dul a t i onT 97 0090 T fel imi hitid %1 bng liisdi nalMty[e%ru p t u re s Q f roe M on s i dera t 16n S for$t hs Pres suri ze ri Surge Ri ne d by X fol l owi ng sthe i gui del i nesiof4 Generi cWetter > 87;lli "Relaxatioisin!Arbitrar Ihe~dssfgiiappF6ach~wa's/"to"                  design"each'c6lhionent~5hd cbmp'onhnt" sup worst break type and location.
The constant forcing function which was used in the design of pipe restraints utilized a force equal to one times the operating pressure times the cross-sectional area of the break.           In addition, the restraints for the 36 inch lines were analyzed by the energy balance technique where the gap was taken into account using the constant forcing function. The RCPs were designed using a static analysis technique with a dynamic load factor of 1.7 times the constant forcing function. All pipes whose ruptures could cause deleterious effects on the other components or piping have been adequately restrained.
4.2.7 MISSILE PROTECTION The major components including reactor vessel, RCS piping, RCPs, steam generators, and the pressurizer are located within three shielded cubicles. Each of two cubicles contain one steam generator, two RCPs, and associated RCS piping.
One of the cubicles also contains the pressurizer. The reactor vessel is located within the third cubicle or primary shield. The reactor vessel head and control rod drives extend into the fuel transfer canal.
4-38                                  (REMXX).

                , 14.2.2.5.11 Reactor Building Subcompartments Pressure Response (continued)
To simulate the physical model and to take advantage of the computer code, the entire system was conveniently divided into several lumped nodes. Each volume was given an average height depending on its lowest and top-most elevations in the actual system. An average cross-sectional area of each node was calculated from existing layouts and/or structures.
The control volumes were connected by two principal types of paths, i.e., surge line (pipe and critical flow check) and a leak path (flow versus area). The flow direction of each path was specified according to the physical understanding of the system behavior. When flow direction is uncertain, the program has bi-directional capability. Details of the mock-up of the whole system by control volumes (nodes) and the paths connecting each is shown in Figures 14-55 through 14-58. The control volumes (nodes) shown on these figures were selected because they represent the actual physical configuration w; thin the containment building.
The break locations in the control volumes do not affect C pressure response, and hence were chosen to be in the mid-plane of the pipe run.
To complement the above program application, several assumptions had to be considered in this analysis:
: a. Although the Flash-2 code handles two phase mixtures (where in fact there is steam-air-water), a modification was made to the code to simulate the initial conditions as a three component atmosphere.
Subsequently, during the transient a two phase mixture is considered.
: b. No heat transfer to the containment walls, floors, or equipment was assumed.
: c. Other assumptions remain intact as described in the Flash-2 report.
The highest calculated steam generator compartment pressures occur for a rupture of the hot leg pipe with an area equivalent to a double-ended break. This pipe is the largest in the RCS, and its rupture results in the highest mass and energy release rates. The blowdown rates were calculated using Moody correlation, with a Co - 1.0, as explained in BAW-10030, " Craft Descript'on of Model for Equilibrium LOCA Analysis Program," October 1971.
Jet Imoinaement Forces:
The forces on targets in the path of escaping fluids are dependent upon the size and shape of the target and its distance from the rupture area. Since the actual shape of the rupture dictates the flow field shape being generated, it is assumed that a typical rupture is circular and the free _ stream. expans. ion o.f _the _ jet conical with an included angle of 30*. An:excepti.onftolthistis:the. Pressurizer SUiis{l~iWE f6EwhithTthe ?jst?[i siconi c alNi tManEincl uded tangi s ;on 20 $as docume.nt ed ; i nj F PCi C al,cul _at i o n; M_97 - 00.91, Calculations of the force on an object is determined by assuming that the dynamic pressure (di - %C (V - V,)') developed at the rupture exit is for the maximum 3
mass flow and pressure conditions. Also, it is assumed that the pressure is inversely oroportional to the cross-sectional area of the conical expansion being generated.          The dynamic pressure used for these calculations was determined assuming V, - 0.0 to provide the maximum dynamic affects since this velocity is dependent on the break conditions.                                                . .   ..
14-69                   {Re_vp_XX).
i

FLORIDA POWER CORPORATION                     !
l                         CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302/ LICENSE NUMBER DPR-72 3F0997-01 i
ATTACHMENT C                                    l List of Regulatory Commitments l
2 4
i i
)
4

U. S. Nuclear Regulatory Comission
           .                                                                  Attachment C 3F0997-01                                                          Page 1 of 1 The following table identifies those actions comitted to by Florida- Power
:            Corporation in this' document. Any other actions discussed in the submittal represents intended or planned actions by Florida Power Corporation. They are described to the NRC for the NRC's information and are not regulatory comitments. Please notify the Manager, Nuclear Licensing of any questions regarding this document or any associated regulatory comitments.
ID NUMBER-                     COMMITMENT             IMPLEMENTATION 3F0997-01-1     FPC will revise FSAR Section 4.2.6.6,  FSAR Revision 24 "LOCA Restraints," to state that December 8, 1997 longitudinal ruptures were eliminated by following the guidelines of Generic Letter 87-11.
3F0997-01-2     FPC will revise FSAR 14.2.2.5.11,      FSAR Revision 24
                                " Reactor Building Subcompartments     December 8, 1997
.                               Pressure Response," to state that the fluid jet created by a circumferential 1
rupture is limited to a 20' included angle rather than 30' as currently described in the FSAR.
4 i
4 1-}}