3F0589-20, Forwards Addl Info Re Bases for Exemption Request for Schedular Relief from Requirements of GDC 4 Re High Energy Line Break.Design Work on 10 Break Locations,Including Jet Shield & Pipe Whip Restraints,Initiated

Forwards Addl Info Re Bases for Exemption Request for Schedular Relief from Requirements of GDC 4 Re High Energy Line Break.Design Work on 10 Break Locations,Including Jet Shield & Pipe Whip Restraints,Initiated

ML20247M697
Person / Time
Site:	Crystal River
Issue date:	05/24/1989
From:	Widell R FLORIDA POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
3F0589-20, 3F589-20, NUDOCS 8906050106
Download: ML20247M697 (19)

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e,e Florida Power C O R POR ATION May 24, 1989 3F0589-20 U. S. Nuclear Regulatory Commission Attention: Document Control Desk

' Washington, D.C. 20555

Subject:

Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 High Energy Line Break

Dear Sir:

Florida Power Corporation (FPC) has had discussions with the staff

'regarding the bases provided in our December 16, 1989 exemption requestLfor schedular relief from the requirements of GDC-4. In particular, the staff requested FPC to elaborate on certain technical details of our undue risk considerations. FPC has reviewed the basis provided and has included additional information where appropriate.

Attachment i reflects the original exemption request supplemented with this information.

FPC has started design vork on 10 of the break locations. The work includes both jet shield and pipe whip restraints. These 10 locations

.were selected as they were the same under both the old criteria [GAI Report 1811) and new criteria (submitted on March 31, 1989]. In I addition, these break locations included those having the most potential for damage. These locations include 4 main steam line breaks, 2 main feedwater line breaks, 2 steam generator blowdown line breaks, and 2 auxiliary steam line breaks (branch lines off the main steam line to the emergency feed. water pump turbine). The main steam and main feedwater breaks have the largest potential for energy release in the event of a' break. These breaks already have some existing shielding and pipe whip restraints. Attachment 3 shows a complete tabulation of all breaks postulated using the revised criteria.

8906050106 890524 gDR ADOCK0500ggg2 goof

' s POST OFFICE BOX 219

• CRYSTAL RIVER, FLORIDA 326294219 + (904) 795-6486 A Florida Progress Company

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May 24, 1989 3F0589-20 Page 2 The schedule provided in the exemption request was based on the information available at the' time of the submittal. The resolution of concerns with the criteria and the experience gained through the work  ;

on the 10 break locations will allow development of a firm schedule. I We anticipate completion in 1990 depending on the need for and scheduling of outages during the next cycle of operation.

Sincerely, Rolf C. Widell, Director Nuclear Operations Site Support RCW/JWT/sdr Attachment xc: Regional Administrator, Region II Senior Resident Inspector

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May 20, 1989

~3F0589-20 ATTACHMENT 1 BASES FOR SPECIFIC EXEMPTION NRC. regulations in 10CFR50.12(a) provide.the requirements for seeking specific exemptions. Under Section 50.12, an exemption may be. granted if the' exemption will not present an undue risk to the public health and. safety and~special circumstances are present. FPC believes the standards for an exemption are satisfied by considering the following:

1. No Undue Risk L The proposed schedular exemption will present no undue risk to public' health and safety because:

a. The principle safety systems have existed since original plant construction and the implementation.of GAI #1811.

Therefore, HELB interactions are'quite likely to have been properly _'dispositioned. We are unable to locate sufficient documentation to rely upon.this as permanent resolution and are, therefore, reestablishing our demonstration of compliance with a more appropriate HELB design basis. Even thoughLFPC failed to consider explicitly HELB effects,.many other criteria (electrical separation, Appendix R, etc.)

provide some degree of protection by reducing the likelihood that any one break can cause loss of safety function by impacting multiple trains.

FPC has not identified any information which would reduce our confidence that the original HELB work was adequately accomplished. Nevertheless, FPC will improve the level of demonstrated compliance as we implement the revised criteria,

b. FPC has examined the pipe stress analysis in the piping systems covered by the HELB criteria to get an understanding of the actual stress levels in the piping compared to the specified allowables at the postulated terminal end break locations. The results of that examination are shown in Attachment 2 (revised). The data shows that postulated break locations are not highly stressed. The reduced stress levels ,

in the. piping systems give FPC further assurance that a terminal end break is a very low probability event.

The staff requested FPC to demonstrate that calculated stress levels using the original design codes (generally ANSI

'B31.1). envelope those calculated using more current codes (ASME III, NC) which form the basis for current staff positions on AIB's, etc. Attachment 4 is a portion of that review. This review continues to demonstrate very low stress levels and hence low break probability.

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e ,May-20,.1989 "U 3F0589-20 l

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c. CR-3 has functional redundancy in the High Pressure Injection

'(HPI) System and the Emergency ' Feedwater (EF) System. That' is, the plant can be' safely shutdown;with continued decay 3 heat removal- in' Hot Shutdown conditions, with secondary heat removal (EF) or primary feed-and-bleed-(HPI). Without-t completing a full' scale systems interaction analysis, it can be concluded.that one of these two systems will be available to'aitigate a break in'aither of the two buildings'in which HELBs are postulated. No HPI System components are located in the Intermediate Building (IB). Consequently, no HELB event in the IB can affect the HPI System performance.

There are EFW System components located within the Auxiliary Building (AB).- FPC has evaluated.the location of this equipment.and. potential HELBs to determine if any adverse interactions are. evident. Based on our evaluation, FPC'can conclude that the EFW System performance will not be affected by a HELB in the AB.

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The staff requested FPC to expand this consideration in at least=two areas: (1) retention of RCS inventory maintenance capability and-(2) possible remote or local operability should automatic function be lost. Attachment 6 is a further discussion of Makeup System reliability.

The survivability of make-up, letdown., MFW isolation, EFW, RB cooling, and RC drain tank isolation capability were more exhaustively evaluated in the Appendix R reviews. These reviews postulated complete loss of all make-up equipment in each of'8 fire zones in the Auxiliary Building and 3 fire zones in the Turbine Building concurrent with removing from service each of~the three make-up pumps. The review demonstrated safe shutdown capability in each case. There are only 12 breaks in the Auxiliary Building. None of these breaks are likely to cause damage in excess of the Appendix R postulation. One break is even less likely to damage all equipment in a zone than would a fire. The Appendix R analysis is, therefore, generally considered bounding and sufficient to demonstrate retained safe shutdown capability even with unprotected interactions. The means utilized in achieving this level of function survivability was the verification of. existing features or modifications to provide a sufficient degree of compartmentalization, such as spatial separation or barriers. FPC chose not to rely on post-fire maintenance, operator action, or exemption. The largest modification effort was directed at separating A and B trains of shutdown systems'by floor, i.e. most A train equipment L powered and controlled via circuits on the 119' elevation and B train circuits cn1 the 958 elevation. Such degree of compartmentalization will assure survivability of safe shutdown systems to function.

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h May 20,.1989 3F0589-20

.WithLregard to remote manual, local, or manual capability.

Esurviving a HELB where automatic function is lost, FPC provides.the following additional support. -The majority of added " targets":were associated with the addition'of an automatic capability to existing manual capability. . Since.

.there1were-approximately=10,000 potentia) targets and several hundred: modification packages.which may have added such-equipment,.it has not been possible to draw any formal conclusions regarding numbers or percentages added, contrasted with pre-existing equipment.. Even though actual

g. quantitative values are not-readily retrievable, the largest modifications-(EFIC,' Remote' Shutdown, Post Accident:

Monitoring, etc.).do not generally impact local controls.

Those should remain protected. Further, Appendix R reviews described above'do' demonstrate a large degree of

. survivability with normal shutdown, even though automatic HELB mitigating systems were not included in this review.

As noted above,.there are relatively few break locations in the Auxiliary-Building ~. This is due, in part,Jto the fact that.one lif the-high energy lines is~the auxiliary steam line that provides steam heating to'the reactor coolant waste evaporators. That line has remained under.the original.

design. criteria (Report #1811]'for high energylline break.

As.such, safety-related equipment in the vicinity of that line must1be protected from. postulated leak cracks at.the

-most adverse location. As a part of reviewing the design basis for CR-3 for HELB,.FPC determined that the. automatic isolation circuitry for the auxiliary steam line that would protect-the environment of the Auxiliary Building would-only work for breaks,-not cracks. To justify continued operation, FPC decided to keep this'line' closed at all times until a permanent fix could be installed. Therefore, this part of the HELB concern at CR-3 has been eliminated until the

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permanent fix is installed. The permanent fix will consider

.all the HELB effects, i.e., jet impingement and environmental r . qualification.

d. CR-3 is located in an area which ja considered to be r ' seismically inactive. The CR-3 p?. ping is designed for approximately 4 times the expected ground motion (0.03g based on current seismic hazards methodology). This fact combined with the historical-perspective of earthquake activity discussed in FSAR Section 2.5.4, Seismology, leads FPC to conclude that the seismic load contribution to the potential for HELBs is significantly overstated.

e .- Increased attention given to the integrity of feedwater piping systems brought about by the Surry event and the actions required by Bulletin 87-01, Thinning of Pipe Walls in Nuclear Power Plants, reduces the probability of a HELB.

FPC's program which is described in our letter dated 3

1

6 May 20, 1989 3F0589-20 September-3, 1987 began with the last refueling outage and is ongoing. As stated in our bulletin response, FPC's program provides additional assurance that adequate structural integrity remains in our high energy feedwater piping systems and CR-3 can continue to operate safely.

f. Licensed operators are trained to mitigate accidents initiated by HELBs. Scenarios are routinely run during simulator training providing opportunities for operators to diagnose conditions caused by HELB and mitigate resulting

. transients. Emergency / Abnormal operating procedures address

-compensatory measures which can be taken in the event that certain equipment fails to actuate or does not perform its intended function. These procedures and training provide added assurance that the plant can be safely shutdown in the event of HELB. Attachment 5 provides a list of instrumentation that could be used to indicate a HELB in a particular system.

For these reasons, FPC believes that adequate protection of the public health and safety will be maintained during the period relief will be in effect.

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t May 20,.1989 3F0589-20 MTKHENT 2 POSTUIATED TERMINAL END FIPE BREAK STRESS IEVEIS AND BREAK THRESH 0ID IJMITS MATERIAL AND STRESS TABLE SLMEARY SYSTEM (S) MATERIAL PIPING ALIDWABLES BREAK THRESEDID Sh SA .8 SA .8 (Sh+SA)

E ,0ISG,FW,EEW,AS A106 GR B 15000 22500 18000 30000 E-1 & W -2 A376TP304 18062 27952 22362 36811

~ MU-6 THRU -10 -A312/27.9 .17500 27875 22300 36300 E-4,-5,-12, & -13 A312/26.9 14000 26875 21500- 32700 MU-14 THRU 17 A312/27.0 18062 27013 21610 36060 E-11 A312/28.2 18062 28200 22560 37010

• • RIGCROUS ANALYSIS **  % OF DATA IN BREAK PIPE PRIMARY SEC'RY- 70IAL THRESHOLD DECENDING NLMBER HAT'L

• SIRESS STRESS STRESS BREAK CRDER STRESS *************

MAIN STEAM SYSTEM (TABLE C-2-1)

MS-1 A106 GR B 5729 4915 10644 35.5 41.2 E -12 MS-5 A106 GR B 5445 752 6197 20.7 35.5 MS-1 E-9 A106 GR B 5934 4233 10167 33.9 33.9 MS-9 E-12 A106 GR B 6797 5550 12347 41.2 27.7 MS-16 MS-15 A106 GR B 8257 0 8257 27.5 27.5 MS-15 E -16 A106 GR B 8296 0 8296 -27.7 20.7 E-5 MS-17 A106 GR B 2733 0 2733 9.1 15.2 MS-26 E-18 A106 GR B 2760 0 2760 9.2 15.0 E -37 MS-19 A106 GR B 2711 0 2711 9.0 14.9 MS-36 MS-20 A106 GR B 2867 0 2867 9.6 14.1 MS-23 E-23 A106 GR B 4234 0 4234 14.1 13.7 MS-29 MS-24 A106 GR B 3917 0 3917 13.1 13.7 E-25 E-25 A106 GR B 4117 0 4117 13.7 13.1 MS-24 E -26 A106 GR B 4570 0 4570 15.2 12.8 E -32 MS-29 A106 N B 4105 0 4105 13.7 11.9 MS-35 MS-30 A106 GR B 3351- 0 3351 11.2 11.4 E -38 E -31 A106 GR B 3246 0 3246 10.8 11.2 MS-30 E -32 A106 GR B 3852 0 3852 12.8 10.8 E -31 MS-35 A106 GR B 3559 0 3559 11.9 9.6 E -20 E -36 A106 GR B 4480 0 4480 14.9 9.2 MS-18 MS-37 A106 GR B 4509 0 4509 15.0 9.1 MS-17 MS-38 A106 GR B 3414 0 3414 11.4 9.0 E -19 I

OISG BLOWDOWN SYSTEM (TABIE C-2-1)

MS-41 A106 GR B 10358 14664 25022 83.4 99.1 E -42 E-42 A106 GR B 15051 14664 29715 99.1 83.4 MS-41

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'3F0589-20 -l

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• • RIGCEDUS ANALYSIS ** '% OF DATA IN IEEAK PIPE PRIMARY SEC'RY TUIAL THRESHOID IECENDING NLMBER . MAT'L

• STRESS STRESS STRESS IEEAK OEEER STRESS *************

FEE [ MATER SYSTEM (TABLE C-2-2)

FW-1 ' A106 GR B 5903 1635 7538 25.1 84.7 FW-28B L, 'FW-6 A106 GR B 9421 867 10288 34.3- 60.7 FW-28A FW-10. A106 GR B 11846 5352 17198- 57.3 57.3 FW-10 FW-14 A106 GR B 5235 1700 6935 23.1 41.7 FW-31 FW-20 A106 GR B'. 10506 601 11107 37.0. 37.0 FW-20' FW-24 'A106 GR B'- 6942 -1664 8606' 28.7 -34.3 FW-6 FW-28A-A106 GR B- ~5771 10931- 16702' 60.7 33.9 FW-32 FW-28B A106 Gt B 6928. 15245 -22173 84.7 28.7 FW-24 FW-28C A106 GR B. 7757 157 7914 26.4 26.4 FW-28C-FW-29 A106 GR B 7459 .103 7562 25.2 25.2 FW-29 1FW-30 A106 GR B- 4899 125. 5024 16.7 25.1 FW-1 FW-31 A106 GR B 4758 7509 -12267 41.7 24.9 FW FW-31 A106'GR B 6753 0 6753 22.5 23.1 FW-14 FW-32 A106 GR B 7480 0 .7480 24.9 22.5 FW-31 FN-32 A106 GR B 6867 3296 10163 33.9 16.7 FW-30 MAKEUP AND PURIFICATION SYSTEM (TABLE C-2-3)

' W-1 A376TP304 5505' 754= 6259 17.0 86.0 MJ-16 E-2 -A376TP304 5373 1097 6470 17.6 62.7 MJ-5

' W A312/26.9 8151 1666 9817 30.0 57.7 MJ-10

. E-5 A312/26.9 14185 6330 20515 62.7 35.5 MJ-8 MJ-6 .A312/27.9 4349 1475- 5824 16.0 31.2 E-13.

MJ-7; A312/27.9 5682 1281 6963 19.2 30.0 MJ-4 MJ-8 A312/27.9 6418 6481 -12899 35.5. 29.3 MJ-12 MJ-9 A312/27.9 6860 151. 7011 19.3 28.5' KI-17 MJ-10 A312/27.9 3507 12862 16369 57.7 24.3 KI-15 MJ-11 A312/28.2 5482 492 5974 16.1 23.4 MJ-14 MJ-11 7A312/28.2- 3949 169 4118 11.1 19.3 MJ-9

. MJ-11 ' A312/28.2 4517- 749 5266 14.2 19.2 MJ-7 MJ-12 A312/26.9 7602 1981 9583 29.3 .17.6- MJ-2 M1-13 A312/26.9 8237 1981 10218 31.2' 17.0' MJ-1

MJ-14.- A312/27.0 4458 615 5073 14.1 16.1 MJ-11 MJ-14 A312/27.0 8447 0 8447 23.4 16.0 MJ-6 Mi-15 A312/27.0 4058 4698 8756 24.3 14.2 MJ-11 MJ-16 A312/27.0 4918 18587 23505 86.0 14.1 MJ-14 MJ-17 A312/27.0 5928 4353 10281 28.5 11.1 MJ-11 EMERGENCY FEEDWATER SYSTEM (TABIE C-2-5)

EF-11 A106 GR B 4859 2 4861 16.2 18.8 EF-13 EF-13 A106 GR B 5650 0 5650 18.8 16.2 EF-11

LMay 20,-1989-

. 3FT)589-20

• • RIGCROUS ANMXSIS **  % OF DMA B BREAK PIPE PRIMME . SEC'RY TUIAL THRESICID IECENDING NLMBER MAT'L

• SIRESS STRESS STRESS BREAK OREER STRESS *************

AUXILIARY STEAM SYSTEM (TABIE C-2-4)

AS-1 . A106 GR B 3355 5798 9153 32.2 42.4 AS-5 AS-1 A106 GR B 6430 4065 10495 35.0. 35.0 -AS-1 AS-1 A106 GR B 4841 9 4850 16.2 32.2~ AS-1~

AS-1 A106 GR B 6724' 418 7142 23.8 28.0 AS-12 AS-5 A106 GR B 5589 58 5647 18.8 26.7 AS-4 AS-5 A106 GR B 10175 2559 12734 42.4 26.0 AS-4 AS-4 -A106 GR B 6435 1589 8024 26.7 24.0 AS-13 AS-4 A106 GR B 7591 210 7801 26.0 23.8 AS-1 AS-10 A106 GR B 6561 473 7034 23.4 23.4 AS-10 AS-12 A106 GR B 5454 2944 8398 28.0 18.8 AS AS-13 -A106 GR B- 7188' O 7188 24.0 16.2 AS-1

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May 20, 1989

- 3 F0589-2 0 ATTACHMENT 3 HELB BREAKS A review of postulated break locations was performed to determine the extent of HELB protective devices installed during original plant construction and subsequent modifications and upgrades. The table below summarizes the review findings.

1. Breaks # Breaks Breaks With Line # # Breaks Partial Completely No Potential Intent System Size Breaks Restrained Shielded Shielded Tarcet_,

MS 24" 4a 4 4 0 0 MS 1-1/2" 9 0 0 0 5b OTSG BD 3" 2 2 0 2 0 AS 6" 3c 3 2 0 0 AS 4" 1 0 0 0 0 EF 6" 2 0 0 0 0 FW 18" 2a 2 1 0 0 FW 6" 7 0 1 0 0 MU 2-1/2" 6 0 0 0 0 MU 3" 4 0 0 0 0 MU 2 0 0 0 0 0 Total 42 11 8 2 5 Footnotes:

a. Design of required jet shields to completely mitigate breaks is in process.

-b. 5 breaks in 1-1/2" MS branch lines will blowdown and whip into the IB floor approximately 2 ft. away.

c. Design of jet shields to completely mitigate 2 of the 3 breaks is in process.

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May 20, 1989 3F0589-20 In conclusion, there are 42 postulated HELB locations outside

-containment, 7 breaks are either shielded or have'no significant jet impingement or pipe whip consequences, 11 breaks are restrained for pipe whip (including all postulated breaks in the 24' MS, 18" FW, 6"

'OTSG blowdown lines),-and 8 postulated breaks are partially shielded by existing jet shield devices.

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I May 20, 1989 i 3F0589-20 f 1

l l ATTACHMENT 4 PIPING ANALYSIS METHODS I

{

i The design basis of piping at CR-3 which is not reactor coolant piping is the USAS Code 331.1 (1967). As part of the review for the I development of the updated HELB criteria, a comparison of the basis behind B31.1 and the more recent ASME Code Section III (1974) was performed. The basis of each Code and a comparison of ASME (1974)

I versus B31.1 (1967) methods for 10 High Energy Piping Analysis problems are discussed in the following paragraphs.

BASIS OF B31.1 (1967)

The fatigue piping component tests of A.R.C. Markl in the 1950's were used as a basis to develop stress intensification factors (SIF's) for B31.1 components. The fatigue tests used a girth buttweld as the baseline component, with other piping components (i.e., elbows, tees, etc.) fatigue lives scaled to the buttweld baseline. The SIF represented " peak stresses" or " secondary stresses", which are considered to be self-limiting stresses from a code viewpoint.

Based on these tests, bending moment loadings for longitudinal thermal expansion stresses were intensified (SIF applied); however, torsional moments were not. Longitudinal pressure stresses were calculated based on classical pipe stress methods. There was no distinction between primary and secondary stresses in this Code (1967).

BASIS OF ASME III (1974)

In 1974, the ASME Code,Section III was released, which departed somewhat from the previous stress calculation methods of the 1971 ASME Code (which parallelled B31.1). The major changes included:

1) Update of the SIF tables to reflect better fatigue data for some piping components such as welds, reducers, and branches.

2) Adopted intensification of the resultant moment (square root-sum-of-the- squares) as the basis for piping stress.

3) Code now discriminates between primary (0.75 i) and secondary (i) stresses in applying the SIF.

4) Other minor changes such as use of mean-wall fiber stress, and defining upset loading allowables (primary) to 1.2 Sh.

l Some of these changes would result in higher calculated stresses (items 1 and 2); and some would result in lower calculated stresses (3 i and 4). A direct comparison was required to assess the overall effect on break postulation, since the break threshold stress was also increased from .8 (Sh+Sa) for B31.1 to .8 (1.2 Sh+Sa) for ASME (1974).

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May 20, 1989

.3F0589-20 Longitudinal pressure stress calculational methods were unchanged from the B31.1 code.

COMPARISON OF RESULTS In order to assess the real significance of the ASME Section III Code (1974) changes as compared to B31.1. (1967) , ten (10) high energy line analysis problems were reviewed. This review included (4) main steam, (2) feedwater, and (4) decay heat system problems, with piping sizes ranging from'24 inch schedule 60 to 6 inch schedule 80, including 48 individual piping components. The 48 components selected represented identified crack locations (i.e., highest stress), and included an evaluation of'ASME (1974) stresses (based on intensification of resultant moment) which were then compared to the B31.1 reported stresses from the existing piping analysis packages.

The results show:

. No new break locations when using ASME Section'III (1974) methods.

. Some piping has been qualified to ASME (1974) and then reconciled

'to B31.1. Breaks / cracks were postulated on reported stress levels for-these analysis problems.

. Of 48 components, only 3 had total ASME primary plus secondary stresses larger than the existing B31.1 reported stresses.

Minimum ASME based break margins of at least 20% (i.e., maximum primary plus' secondary stress / break threshold ratios no larger than 0.80)were identified, most below 50%. The ASME calculated stresses did not cause break threshold margins to increase by more than 10%.

. Even though B31.1 methods intensified only bending moments, the inclusion of the 0.751 (primary stress) requirement of ASME (1974) resulted in generally lower ASME (1974) stresses than B31.1 values. The increase break threshold also contributed to overall higher break threshold margins for ASME based break analysis methods.

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.May 20, 1989 3F0589-20 1

l ATTACHMENT.5 INSTRUMENTATION TO DETECT A HELB SYSTEM: MAKE-UP BREAK # ALERTING CONDITIONS MU-1,2 1) Decreasing Letdown Flow MCB Indication (MU FI)

2) Dec. Letdown Temp MCB Indication (MU TI)-

3) Dec. Prefilter delta p MCB Indication (MU-81-DPI)

4) Dec. Letdown Radiation MCB Indication / Alm.

(RM-L1)

5) Dec. Pressurizer Lvl. MCB Ind./ Alm. (RC LI1, 2) (RC-1-LS2 )

6) Dec. Filter delta p MCB Indication (MU-18-DPI)

7) Dec. Makeup Tank Lvl. MCB Ind./ Alm. (MU LIR1) (MU-14-LS1)

8) Dec. Makeup Tank Press. MCB Ind./ Alm. (MU PIR1) (MU-17-PS)

MU-5, 6, 7, 8, 1) Dec. Makeup Flow MCB Indication (MU-24-FI).

9, 10, 11, 15, 2) Dec. Total Seal Inj. Flow MCB Ind./ Alm. (MU FI) (MU-27-FS) 16, 17

3) Dec. RCP Seal Inj. Flow MCB Ind./ Alm, (MU FI1, 2, 3, 4 3) (MU FS1, 2, 3, 4)

4) Dec. Makeup Tank Lvl. MCB Ind./ Alm. (MU LIR1) (MU-14-LS1)

5) Dec. Makeup Tank Press. MCB Ind./ Alm. (MU PIR1) (MU-17-PS)

6) Dec. Pressurizer Lvl. MCB Ind./ Alm. (RC LI1, 2) (RC-1-LS2) i f

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May 20, 1989 3F0589-20 SYSTEM: MAIN STEAM POSTULATED BREAK ALERTING CONDITION' MS-1 1. Depressurization of OTSG A - as sensed by MS-106-PT, MS-107-PT, MS-108-PT, and MS-109-PT

2. Initiation and alarm of MSLI

3. Automatic closure and alarm of MFWLI

4. Start and alarm of both EFW pumps

5. EFW Vector Logic isolation of bad'OTSG MS-5 Same as MS-1 MS-9 1. Depressurization of OTSG B - as sensed by MS-110-PT, MS-111-PT, MS-112-PT, and MS--

113-PT

2. Initiation and alarm of MSLI

3. Automatic closure and alarm of MSWLI

4. ' Start and alarm of both EFW pumps

5. EFW Vector Logic isolation of bad OTSG MS-12 Same as MS-9 MS-41 1. Depressurigation of OTSG A

2. Initiation of both EFW pumps on low level in OTSG A MS-42 1. Depressurization of OTSG B

2. Initiation of both EFW pumps on low level in OTSG B MS-45 Same as MS-9 MS-46 Same as MS-1 MS-47 Same as MS-1 MS-48 Same as MS-9 ,

i MS-49 Same as MS-9 MS-50 Same as MS-1 MS-51 Same as MS-1 MS-52 & -53 Same as MS-1 1

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i May 20, 1989 3F0589 SYSTEM: AUXILIARY STEAM POSTULATED BREAK ALERTING CONDITION AS-1 1. Depressurization of OTSG B - as sensed by MS-110-PT through MS-113-PT

2. Initiation and alarm of MSLI

3. Automatic closure and alarm of MFWI

4. . Start and alarm of both EFW pump

5. Vector Logic isolation of bad OTSG AS-5 1. Depressurization of OTSG A - as sensed by MS-106-PT through MS-109-PT

2. Initiation and, alarm of MSLI

3. Automatic closure and alarm of MFWI

4. . Start and alarm of both EFW pumps

5. EFW Vector Logic isolation of bad OTSG AS-4 1. Depressurization of OTSG A & OTSG B - as sensed by MS-106-PT through MS-113-PT

2. Initiation and alarm of MSLI

3. Automatic closure and alarm of MFWI

4. Start and alarm of motor driven EF pump

5. Alarm that turbine driven EF pump failed to start AS-12 Same as AS-4 12

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May 20, 1989.  !

'3F0589 l 1

SYSTEM: MAIN FEEDWATER/ EMERGENCY FEEDWATER - S.G. "B" BREAK ALERTING CONDITIONS FW-14 1. OTSG pressure decrease (MS-110, 111, 112, 113-PT)

2. OTSG level decrease (SP-29, 30, 31, 32-LT)

3. Initiate EFW & alarm of both EFW pumps started

-4. Increase flow (SP-8B-FI1, FI2)

5. High RCS p essure trip

6. Turbine trip EF-13 1. OTSG pressure decrease-(same as FW-14)

2. Main FW & MS isolation

3. High EFW flow to good OTSG

4. Alry.a of both EFW pumps started

5. EFW Vector Logic isolated bad generator FW-20 1. OTSG pressure increase (same as FW-14) 2.- OTSG level decrease (same as FW-14)_

3. Initiate EFW & alarm of both EFW pumps started

4. Increase flow (SP-7B-FI1) (SP-12B-DPT)

5. Decrease flow (SP-8B-FI1, FI2)

6. High RCS pressure trip

~7.- Turbine trip FW-24, FW-32 1. OTSG pressure increase (same as FW-14)

2. OTSG 1evel decrease (same as FW-14)

3. Initiate EFW & alarm of both EFW pumps started

4. Decrease flow (same as FW-20, conditions 4 &

5)

5. High RCS pressure trip

6. Turbine trip l-13 l

i May.20, 1989

3F0589-20 SYSTEM: MAIN FEEDWATER/ EMERGENCY FEEDWATER - S.G. "A" l

BREAK ALERTING CONDITIONS FW-1 1. OTSG pressure decrease (MS-106, 107, 108, 109- .

PT)

2. OTSG 1evel decrease (SP-25, 26, 27, 28-LT)

3. Initiate EFW & alarm of both EFW pumps started  !

4. Increase flow (SP-8A-FI1, FI2)

5. High RCS pressure trip

6. Turbine trip EF-11 1. OTSG pressure decrease (same as FW-1) i

2. Main FW & MS isolation

3. High EFW flow to good OTSG

4. Alarm of both EFW pumps started

5. EFW Vector Logic isolated bad generator FW-6, FW-28 1. OTSG pressure increase (same as FW-1)

2. OTSG level decrease (same as FW-1)  !

3. Initiate EFW & alarm of both EFW pumps started

4. Increase flow (SP-7A-FI1) (SP-12A-DPT)  ;

5. Decrease flow (SP-8A-FI1, FI2) l

6. High RCS pressure trip '

7. Turbine trip i FW-10, FW-31 1. OTSG pressure increase (same as FW-1)

2. OTSG 1evel decrease (same as FW-1)

3. Initiate EFW & alarm of both EFW pumps started

4. Decrease flow (same as FW-6, conditions 4 & 5)

5. High RCS pressure trip

6. Turbine trip l Refer to Attachment 6 for a discussion of the reliability of the I Makeup System when the breaks are considered.

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May'20, 1989 3 F0589-2 0..

b- ATTACHMENT 6

~HIGH' ENERGY LINE BREAKS j MAKE-UP SYSTEM RELIABILITY l

The.High Energy Line Break (HELB) analysis performed Crystal 1 River

-Unit 3 identified twelve ~-(12) credible break locations, as described-in Impell Report No.- 03-0920-1186,. dated March 1989. None of the

_ identified breaks can prevent the MU System from performing its inventory control function.

The redundancy built into the Make-Up System is sufficient to overcome any challenge to system function presented by a High Energy Line Break.

Two'(2) independent supply lines ensure borated water is supplied to the Make-Up - Pump -(.MUP) suction header.

A' common suction header supplies three (3), 100% capacity MUP's.

A disabled portion of this header.can be isolated'aither remotely or manually, ensuring two (2) of three (3) MUP's are supplied with

. water.

Theithree (3), 100% capacity MUP's discharge to a common header, allowing-any' single pump to supply multiple injection' lines. This common discharge header contains both manual and remotely operated valves which can isolate any disabled section, ensuring three (3) of five (5) potential flow paths are available to supply the Reactor Coolant' System (RCS) with~ borated water.

There are five (5) potential flow paths for borated water into the RCS - four'(4) High Pressure Injection (HPI) lines, and one (1)

Reactor Coolant Pump seal injection line. E Jh of these five (5) lines is independent and can be placed in service or isolated from the control room.

Keeping the redundance built into the Make-Up System in mind, each

. potential break location has been reviewed to discount any credible threat to system function.

Break MU-1:

Potential break location is downstream of the letdown coolers and can be isolated remotely from the control room. .This break is in the-vicinity of the four (4) HPI injection lines and the four (4) remotely operated valves used to place them in service. The possibility of this break disabling all four lines is considered extremely remote.

- The remote operated valves needed to place an injection line'in service range between 4 and 19 feet from the break.

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t lMay 20, 1989 3F0589-20

.The four (4) remotely operated valves and the potential break are on slightly different elevations ( 2 ft.).

The potential break would have to drastically change its angle of incidence by 120 0 to impinge upon and disable each remotely operated valve.

The four (4) injection line penetrations span 18 ft., precluding the' possibility of pipe whip impacting and disabling all four (4) lines.

Break MU-2:

This potential break location is slightly downstream of break MU-1, just prior to the black orifice on the 119' elevation. This location can potentially impact MU System components on both the 95' and 119' elevation. There are no MU System components on the 119' elevation associated with RCS inventory control. This break cannot totally disable those MU System components associated with RCS inventory control for the same reasons outlined under break MU-1.

Break MU-5:

This potential break is in the normal make-up line, just prior to penetration #435. The short run of pipe available for whip (approximately 4_ft.) is not sufficient to impact other injection lines (two range from 10 to 16 feet away). Remotely operated valves necessary to place injection lines in service range from 12 to 24 feet away, and are on slightly different elevations (12 ft.). The probability _of this impinging jet changing its angle of incidence by 1200 and disabling remotely operated valves at varying distances is extremely remote.

Breaks MU-6, 7. 8. 9. 10:

This potential breaks MU-7, 8, 9, and 10 are located in the South Auxiliary Building on the 95' elevation, with MU-6 immediately above on the 119' elevation. Break MU-6 can be dismissed, since there isn't any equipment associated with the make-up function on that level.

Breaks MU-7, 8, 9, and 10 located upstream and downstream of the reactor coolant pump seal injection filters cannot credibly disable the four (4) HPI injection lines for the following reasons:

- These four (4) breaks are located in a separate room. Any potential jet would have to travel over 20 feet, with an accuracy of i2 0 just to exit the room through a 3 ft. wide door.

- After leaving the room, the impingir.g jet would have to travel an additional 24 ft. just to contact the nearest injection line.

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May 20, 1989 3F0589-20 The impinging jet can contact only three (3) of four (4) remotely operated valves needed to place an injection line in service, the fourth shield by an existing concrete wall.

The impinging jet would have to travel in excess of 60 feet, with an angle of incidence accurate to less than 10, to contact the i furthest remotely operated valve.

Break MU-11:

This potential break occurs on the 95' elevation of the Auxiliary Building, in the make-up valve room. The break occurs at an anchor, which will lessen the impact of pipe whip, if any. The resulting jet, constrained from changing direction, cannot disable the four (4) supply lines which run through the room. Consider that:

Two of the injection lines are 12 and 18 feet from the break, j respectively, and in different directions. A constrained jet would have to change its angle of incidence by 1800 just to contact both lines.

The valve room contains remotely operated valves used to isolate cections of the common discharge header. Even if the impinging jet disabled both remotely operated valves, two (2) manually

g. operated valves are available which serve the same function.

These valves may be operated from outside the make-up valve room via remote handwheels, thus, the operator is not in danger.

Breaks MU-15, 16. 17:

These potential breaks are located at the discharge of each make-up pump. While these pumps are located in a common room on the 95' elevation'of the Auxiliary Building, each is separated by a concrete partition. A break at any pump discharge will not disable the remaining two (2) pumps. In addition, all remotely operated valves required to isolate the suction and discharge of the disabled pump are located outside of the make-up pump room.

There is adequate redundance built into the Make-Up System to ensure continued RCS inventory control following any postulated Make-Up System HELB. A single HELB in the MU System is not capable of disabling multiple injection lines, control valves and pumps - the targets are simply too numerous and diversely located.

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