ML20247H076

Vol 1 to Bit Elimination Study for Diablo Canyon Units 1 & 2

ML20247H076
Person / Time
Site:	Diablo Canyon
Issue date:	11/30/1988
From:	Mchugh C, Petzold J, Wooten L WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML16341F321	List: ML20247H076
References
WCAP-11938, WCAP-11938-V01, WCAP-11938-V1, NUDOCS 8909190210
Download: ML20247H076 (140)

Text

{{#Wiki_filter:-. WESTINGHOUSE CLASS 3 WCAP-11938 VOLUME 1 o O BIT ELIMINATION STUDY FOR DIABLO CANYON UNIT I AND 2 by L. A. Wooten C. J. McHugh J. S. Petzold L. C. Smith 3 A. ii. Sicari November, 1988 3 Approved by : ' (( f ELR M. P. Osborne, Manager Transie Analysis II B. S. Monty, Managtf Containment and Operational Analysis WESTINGHOUSE ELECTRIC CORPORATION Power Systems Division P.O. Box 355 l l. Pittsburgh, Pennsylvania 15230 $k $kfibb0 bbb QS a

1 i LEGAL NOTICE l* This report was prepared by Westinghouse Electric Corporation (H), as an account of work sponsored by the Pacific Gas and Electric Co. Neither Westinghouse, nor any person acting on its behalf either: (a) makes any warranty of representation, express or implied, with respect to the accuracy, completeness, or usefulness of the information contained in this report, or that the use of any information, apparatus, method, or process disclosed in this report may not infringe privately owned rights; or (b) assumes any liabilities with respect to the use of, or for damages resulting from the use of, any information, apparatus, method, or process disclosed.in this report. 6 O 3 P i

e O ABSTRACT This report provides safety information needed to support the removal of the Boron Injection Tank (BIT) from Diablo Canyon Nuclear Power Plants, Units 1 and 2. The analysis covers Standard, Transition, and Vantage 5 fuel loadings. Two different types of steam line protection logic are covered. Results are presented to allow for the reduction of the Low Steam Line Pressure Setpoint. In addition, the possibility of future implementation of digital Eagle 21 hardware for the steam line break logic was considered. The primary transient affected is the Steam Line Break Event. LOCA and other Non-LOCA accidents are not adversely affected by BIT removal. This event is analyzed for three broad subcategories Q of interest: 1) Core Response, 2) Inside Containment Response, and 3) Outside Containment Response. Core Response and Inside Containment response ara discutsed in Volume 1. Outside Containment Response is discussed in Volume 2. Conclusions are reached for Core Response regarding the acceptability of BIT removal. Temperature and Pressure results are provided for the Inside Containment Response. Mass and Energy results are l presented for the Outside Containment Response. Provided the i Equipment Qualification results are acceptable, it is feasible e l to eliminate the BIT at Diablo Canyon Unit I and 2. 6 ii i L 4

Y TABLE OF CONTENTS SECTION PAGE TITLE

---

VOLUME 1 -----

1.0 1-1 Introduction. 2.0 2-1

Background

3.0 3-1 Core Response 4.0 4-1 Inside Containment Response 5.0 5-1 Documentation 6.0 6-1 Summary and Conclusions 7.0 7-1 References APPENDICES c A FSAR Revisions B Technical Specifications

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VOLUME 2 -----

1.0 1-1 Outside Containment Response 2.0 2-1 References APPENDICES A Outside Containment Mass / Energy Data a l I iii l 1 I

1.0 INTRODUCTION

Westinghouse has developed improved analytical techniques which allow a reduction in the Boron Injection Tank (BIT) concentration or removal of the BIT. This report provides background information on the BIT design basis, reasons why tank removal nay be desirable, as well as a summary of analytical results which demonstrate the feasibility of removing the B:T entirely for Diablo Canyon Units 1 and 2. The work performed in the Non-LOCA area covers both units. The results and conclusions presented can be applied directly to the current plant configurations. The analyses and evaluations provide results also intended to apply to the following plant configurations: 1. Standard, Transition, or Vantage 5 fuel loading, 2. The current steam line break logic at the plant (referred to as "Old" ~ Steam line break protection logic), and more recent logic currently employed at other plants (referred to as "New" Steam line break protection logic) which may be employed at the plants in the future. These two types of logic are depicted by Figure 1-1, 3. The current low steam line pressure setpoint, as well as the possibility of lowering the setpoint to as low as 15 psia (or the bottom end of the instrument scale), 4. The current analog equipment at the plant, as well as possible future upgrades to the Eagle 21 digital hardware system. The work performed is somewhat generic since it is intended to apply to all of the above conditions and in any combination for the Diablo Canyon Units. When the above changes are made to the plants in the future, this work is intended 1-1 .I

to still apply, and a confirmation of the continued validity of the results and l conclusions should be all that is required. This report provides the Pacific Gas and Electric Company with all the information essential in evaluating the BIT removal as it relates to the Chapter 15 Non-LOCA accident analysis and Chapter 6 Steam Line Break mass and energy analysis for the Diablo Canyon Updated FSAR (Reference 1). Equipment qualification is not currently within the Westinghouse scope of supply for these units. To complete this work, Pacific Gas and Electric will have to evaluate equipment survivability for both inside and outside containment response using the mass and energy data provided in this report. 'O e D D 1-2

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. Hi. CCNTAI W NT Putt $sJRE v v v v v v v sitAu.tw sArrTv Is0LAT2DN IkKCT3DN 'e SYSTEM l (OLD STEAMLINE BREAK PROTECTION) brP Y@ W.AT# CPcRATIONE) A5 ^k WT WTEAAIT PRES 12EM CCNTal e mss #E Pacss a mssa mssa .1c l L j i V V y V V v srcasi< sArrTv i ISQLATICN INKOTICW L-SYSTEM 2 (NEW STEAMLINE BREAK PROTECTION) Steamline Break Protection Systems FIGURE 1-1 i 1-3 {

2.0 BACKGROUND

Current Design Basis The BIT is a component of the Safety Injection System whose sole function is to provide concentrated boric acid to the reactor coolant to mitigate the consequences of postulated steamline break accidents. Although the BIT acts to mitigate steamline breaks of various siizes occurring from any power level, the cases which serve as the Westinghouse.steamline break licensing basis for Core Response, and which define the existing requirements on the minimum BIT boron concentration, are as follows: For the " hypothetical" steamline break, i.e., double ended rupture of a main steamline, the radiation releases must remain within the requirements of 10CFR Part 100. This is the ANSI NI8.2 criterion for Condition IV events, " Limiting Faults." Westinghouse conservatively meets this for the Diablo Canyon Units by demonstrating that.the DNB design basis is met, the criterion typically used for Condition 11 events. For the " credible" steamline break, (i.e., the failure open of a single steam generator relief, safety, or turbine bypass valve),.the radiation releases must remain within the requirements of 10CFR Part 20. This is the ANSI N18.2 criterion for Condition 11 events,." Faults of Moderate Frequency." Westinghouse has conservatively met this criterion by showing that no return to criticality is achieved, although showing that the DNB design basis is met is traditionally sufficient. For Mass and Energy releases from a steam line break both inside and outside containment, essential equipment must be capable of performing its required function. 2-1

Chapter 6 of the Diablo Canyon FSAR provides the current results for the Inside E Containment mass and energy response. Additionti information is also provided in Chapter 3. The containment structure must be capable of surviving the pressure transient. The criterion used by Westinghouse to demonstrate that this requirement is met is to'show that the peak pressure remains below the Containment Design Pressure of 47 psig, or the integrated leak rate test-pressure (PA of the Technical Specifications, Section 4.6.1) of 47 psig. The . containment and certain essential equipment inside containment must also be capable of performing their required functions given the temperature transients istulated for Inside Steam Line Break accidents. The essential equipment and .ontainment structure have been shown to be capable of surviving the ~ temperature transients experienced due to steam line breaks of various sizes by Westinghouse providing temperature histories for a spectrum of accidents and Pacific Gas and Electric determining that the equipment qualification is acceptable. Results of the Westinghouse calculations are presented in Chapter 6 in the form of temperature plots, and in Chapter 3 of the FSAR in terms of peak Temperature reached. s Technical Specifications In order to assure the validity of the safety analyses performed to verify that the Westinghouse criteria are met, technical specifications are applied to the BIT and associated equipment. Specifically, the Diablo Canyon Technical Specifications, (Reference 2), currently assure that the boric acid concentration is maintained in excess of 20,000 ppm, with a minimum contained borated water volume of 900 gallons, and a minimum solution temperature of-145'F. Heat tracing is necessary to maintain the tank and associated piping at a sufficiently high temperature so that the minimum concentration requirements may be met. Furthermore, the safety-related nature of the boric acid system requires that the heat trace systems be redundant. 2-2

l The required solubility temperature imposes a continuous load on the heaters, ) J and low-temperature alarm actuation and heater burnout have occurred in some operating plants. Violation of the Technical Specification on concentration in ~ the BIT poses availability problems in that recovery is required within a very short time. If the concentration is not restored within one hour, the plant must be taken to the hot standby condition and borated to the equivalent of 1% f AK/K at 200*F. Thus, this requirement has a potentially serious impact j on plant availability. In addition, the high boric acid concentration makes recovery from a spurious safety injection signal (which results in injection of the DIT fluid into the reactor coolant system) time consuming and costly. BIT Removal These potential difficulties unfavorably affecting plant availability, operability, and maintainability can be drastically reduced in severity or eliminated by entirely deleting the BIT system. The effect of BIT removal is discussed in the following sections. BIT removal will be discussed as it relates to 1) Core Response, 2) Inside Containment Response, and 3) Outside Containment Response in three separate sections. Additional information will ~ then be provided on Technical Specification and other documentation changes. c> 2-3

3.0 CORE RESPONSE 9 Introduction . The only Non-LOCA accident analyses which are adversely affected by boron-reduction-or BIT removal are the steamline break transients. This section discusses the impact of BIT removal as it relates to Core Response due to-Steam Line Break. The core response is discussed as it relates to the Credible and Hypothetical Break accidents as analyzed previously with the BIT.in the Diablo Canyon FSAR, Sections 15.2.13,15.3.2, and 15.4.2, (Reference 1). Marked up replacement' sections for these FSAR sections cre provided in Appendix A. Analysis The following cases, which are the ones analyzed in the Diablo Canyon Unit'l b ' and 2 Final Safety Analysis Report, are reanalyzed for the no-BIT system: " Hypothetical" Steamline Break, with and without offsite power available, for the largest double ended rupture of a steam pipe. " Credible" Steamlir.e Break, with offsite power available, for the largest j single failed open steam generator relief, safety or steam dump valve. For the hypothetical break, the same criteria is applied to the BIT removal analysis as is applied in the FSAR (Reference 1). That is, for the most severe l Condition IV break, Westinghouse will show that the radiation releases are within the requirements of 10CFR Part 100 by demonstrating that the DNB design basis is met. The steamline break DBA dose calculations performed for the FSAR use a conservative fuel failure level of 1 percent (Reference 1, Section 15.5.18), although the core analyses shows that no consequential fuel failures are anticipated. 3-1 l-L - _--

1 The credible steamline break analysis will be performed using a new criterion j whereby the plant may return to criticality but no damage may occur to the fuel. This constitutes a relaxation of the conservative internal Westinghouse criterion for Class 11 events used previously for Diablo Canyon, but is an internal Westinghouse criterion which has been used in more recently approved Westinghouse analysis (Eg. Reference 3). This new criterion is in compliance with the criteria used by the NRC and ANS, which require that releases during steamline break accidents remain within the limits set forth in 10CFR Part 20. This limit could be met with a return to criticality if it is assured that there is no consequential fuel damage. As in the Diam Canyon Unit FSAR steamline break analysis, the system transient parameters, i.e., RCS pressure, temperatures, steam flow, core boron concentration and core power are calculated using the LOFTRAN (Reference 4) system transient analysis computer code. This computer code includes models of the reactor. core, steam generators, pressurizer, primary piping, protection systems and engineered safeguards systems. The changes in system volumes, initial conditions, and other design information corresponding to the two units are conservatively accounted for in the analysis. The analysis is set up to conservatively account for a low steam line pressure setpoint of no less than 15 psia (uncertainties must be added to this value in determining the plant setpoint) for either the "new" or "old" steam line break logic. Both "new",and "old" steam line break logic as depicted in Figure 1-1 are conservatively accounted for. The following systems provide the necessary mitigation of an accidental depressurization of the main steam system. i (1) Safety injection system (SIS) actuation from any of the following: 0LD STEAM LINE BREAK LOGIC-(a) Two-out-of-four low pressurizer pressure signals o (b) High differential pressure signals between steam lines. 3-2 w_______-__________.

NEW STEAM LINE BREAK LOGIC-(a) Two-out-of-four low pressurize pressure signals (b) Low steam line pressure in any line (two-out-of-three per line) (2) The overpower reactor trips (neutron flux and aT) and the reactor trip accurring in conjunction with receipt of the safety injection signal. (3) Redundant isolation of the main feedwater lines: Sustained high feedwater flow would cause additional cooldown. Therefore, a safety injection signal will rapidly close all feedwater control valves, trip the main feedwater pumps, and close the backup feedwater isolation valves. j The following functions provide the necessary protection against a steam pipe rupture: (1) SIS actuation from any of the following: 0LD STEAM LINE BREAK LOGIC-(a) Two-out-of-four low pressurizer pressure signals (b) High differential pressure signals between steam lines. (c) High steam line flow in two main steam lines (one-out-of-two per line) in coincidence with either low-low RCS average temperature or low steam line pressure in any two lines s (d) Two-out-of-three high containment pressure. O 3-3 l

NEW STEAM LINE BREAK LOGIC-(a) Two-out-of-four low pressurizer pressure signals (b) Low steam line pressure in any line (two-out-of-three per line) (c) Two-out-of-three high containment pressure. (2) The overpower reactor trips (neutron flux and aT) and the reactor trip occurring in conjunction with receipt of the safety injection signal. (3) Redundant isolation of the main feedwater lines: Sustained high feedwater flow would cause additional cooldown. Therefore, a safety injection signal will rapidly close all feedwater control valves, trip the main feedwater pumps, and close the backup feedwater isolation valves. (4) Trip of the fast acting main steam line isolation valves on: (a) High steam flow in two main steam lines in coincidence with either low-low T r low steam line pressure in any two lines (old avg steam line break protection), or low steam line pressure in one line (new steam line break protection) (b) High-high containment pressure, in the analysis the following values were used as trip setpoints:

1) Low pressurizer pressure SI-1695 psia

2) High differential pressure between steam lines - not used O

?-4 _ _ _ ______ - _ - _____-__________-____-_____s

i

3) High steam line flow coincident with low-low T or low steam line avg l-pressure

_not used in models, values used outside of models are: Power Levgl Setooint (fract nom power) (fract nom flow) 0.0000 0.5727 0.2000 0.5727 0.3143 0.7295 0.4286 0.8509 0.5429 0.9537 0.6572 1.0445 0.7715 1.1266 0.8858 1.2023 1.0000 1.2727 1.2000 1.3862

4) High containment pressure - not used

5) High-High containment pressure - not used o

6) Low steam line pressure - 15 psia

7) Overpower Delta T - (not used)

8) High nuclear flux = 1.12 (not used)

The BIT and associated piping are conservatively modeled as being in place and full of 0 ppm water. The BIT water volume used is 139 ft3. The delay time for delivering the 2300 ppm borated water to the primary system covers a 23 second j pure delay for the cases with power and a 38 second pure delay without power. l 3 P 3-5 i j

4 This is tha delay between when the process parameters being measured reach the l setpoint, to the time when the charging pumps are at full flow. In addition to this delay, there is an additional delay to purge the 0 ppm water between the RWST and the primary system. The modeling of delay times and volumes is consistent with the current Technical Specifications and allows for the BIT to be removed completely, bypassed, or remain in place, so long as the volume filled does not exceed 139 ft3 and the SI pumped flow rate performance remains ) the same. Boron can also be provide by the 2200 ppm accumulators. ) i To cover Vantage 5 fuel and the transition cores, consistent and conservative parameters between Standard fuel and Vantage 5 fuel analyses were used. The 1979 ANS decay heat curve is used in this analysis (Reference 5). Limiting coefficients for the Overpower Delta-T Trip Setpoint are used. The uncertainties used in the analysis for temperature and pressure are limiting. Temperature uncertainty for the full power cases analyzed, (see Low Steam Line Pressure Sensitivity discussion below), included 7.2 degrees F margin. Pressure uncertainty allowed for is 38 psi. Limiting values for both reactor units, both types of steam break protection logic, and both types of fuel are used. As a result a single enveloping set of analyses is performed. o For more details on the methodology used, the protection, and the criteria, see the marked up Chapter 15 Section provided in Appendix A. Results As a point of reference, Figures 3-1 to 3-5 and 3-11 depict the large double ended rupture upstream of the flow restrictor from the Diablo Canyon FSAR, with offsite power available. This is the limiting DNB transient of the FSAR cases analyzed. The plant is initially assumed to be at hot zero power at the O 3-6 l 4 l ___________________________J

minimum required shutdown margin. Following the break, the RCS temperatures and pressures decrease rapidly, and in the presence of a large End of Life (E0L) moderator coefficient of reactivity, the reactor returns critical with the rods inserted, assuming the most reactive RCCA in the fully withdrawn position. The reactor power increases at a decreasing rate until boron from the safety injection system reaches the core and begins to offset the positive reactivity insertion caused by the cooldown. The core is subsequently brought' suberitical with boron injection, aided by the abatement and eventual termination of steam flow from the broken steam generator. Figures 3 6 o 3-10 and 3-12 show the similar transient assuming no offsite power available. The reactor coolant pump coastdown reduces the thermal coupling between the secondary side and the_ primary side, reducing the cooldown experienced by the core. The additional delay for starting the emergency A/C ' diesel generators also accounts for a slightly longer time delay to deliver-the RWST boron to the core. Although preventing clad damage is not necessary for Condition IV events, the analysis shows that the DNB design basis is met, i.e., the DNB Ratio remains greater than the limiting value. Therefore, the dose evaluation, which is_ performed assuming 1 percent failed fuel, continues to demonstrate that the Condition IV accident criteria are satisfied. 1 Figures 3-13 to 3-15 depict transient parameters for the limiting Condition 11 steamline break, assuming no BIT. For the current FSAR case, the -eactivity plot for this case shows that the reactor remains suberitical. Thi.; Westinghouse criterion assures that the DNB design basis is met in a very conservative manner. In order to allow elimination of the BIT, this Westinghouse criterion must be relaxed to allow a subsequent return to power for the Condition 11 transients, but the DNB design basis must be met in order o 3-7 l

'to meet the 10CFR Part 20 dose requirements. Figure 3-15 shows that criticality is attained and sustained for the BIT elimination. DNB analyses for this case shows that the DNB design basis is met and no fuel failures are predicted. This conclusion is also consistent with the conclusion drawn on the Condition IV breaks, since no _ violation of the DNB design basis was calculated for the more. extreme Condition IV, double ended ruptures. "Old" and "New" Protection Logic The. input parameters used were selected to model either form of protection logic in a conservative fashion. Based on the actuation signals obtained, either protection system is covered by the current analyses. The credible breaks rely on the low pressurizer pressure signal for the important actuations. Therefore, either form of logic applies. For the Hypothetical breaks the low steam line pressure setpoint is relied upon for actuation of-SI, and the coincident logic for high steam flow is met. This fulfills the same function for both the old logic that requires coincidence and the new logic that is actuated on low steam line pressure alone, thereby covering both logic types. a Low Steam Line Pressure Sensitivity To address lowering the Low Steam Line Pressure setpoint for either the "new" or "old" steam line break protection logic, hot full power runs were made to assess the sensitivity of the analysis to a setpoint as low as 15 psia. The analysis is consister.t with the methodology reported in WCAP-9226-R1, (Reference 6). The hot zero power cases for the credible breaks as presented in the FSAR and as analyzed here for BIT removal do not actuate the protection - provided by the Low Steam Pressure setpoint, and therefore, they do not need to be rerun for this sensitivity. The hypothetical break cases were analyzed with a low steam line pressure setpoint of 15 psia to allow for the slightly longer time to reach the pressure setpoint. P 3-8 ______J

Note that the 15 psia setpoint is still reached and continues to provide protection'for the large breaks since the indicated steam line pressure-is f modified by a~ lead / lag function prior to being compared to the setpoint'. Since for. large breaks, the rate of pressure drop is fast, the lead / lag anticipates a j very low pressure being achieved, and the Low Steam Pressure setpoint is reached. The hot full power analyscs without the BIT demonstrate that the DNB criterion is satisfied for a spectrum of break sizes ranging from.1 Ft2 to the maximum break size of 5.6 Ft2. Therefore, the Low steam line pressure setpoint may be ' lowered to a value as low as 15 psia without adversely affecting the core response analysis. Digital Equipment As a part of the work performed, the impact of future installation of digital equ!9 ment was investigated. Specifically, the installation of Eagle 21 hardwa m was considered. ~It was determined that steam line break protection logic should not change in relation to the important Non-LOCA parameters. This &ssumes that the design will constitute a " Form Fit and Function" replacement of current analog hardware. Therefore, the analyses performed should continue i - to remain valid if Eagle 21 is implemented. L o 3-9 ~ l

} l i Conclusion I In conclusion, calculations have been performed for the Diablo Canyon Units j ~ which show that, if the internal Westinghouse criterion prohibiting a return to criticality for the credible break is relaxed to the present NRC criterion for Condition II accidents, Pacific Gas and Electric Company may eliminate,... bypass, or dilute the BIT as far as Core Response criteria are concerned. Furthermore, either the "Old" or "New" steam line break protection logic may be used, and a Low Steam Line Pressure Setpoint which is as low as 15 psia is accounted for. O ) i o 3-10

TRANSIENT RESPONSE TO STEAM LINE BREAK AT EXIT OF STEAM GENERATOR WITH WATER INJECTION AND OFFSITE POWER g s00. < =WC < c_ 500. < ew 5E 82 400.< Eit U" 500.< ce 200 2. 50. 100 150. 200. 250. 500. 550. TIME (SEC) s ? 3-11

FIGURE 3-2 DIABLO CANYON TRANSIENT RESPONSE TO STEAM LINE BREAK AT EXIT OF STEAM = GENERATOR WITH WATER INJECTION AND OFFSITE POWER 2 Gb 2EDD. 200D. PRESSURIZER EMPTIES AT 16 SEC w E 1500. 10D0. o O ~ EDD. c e R u< D' E O. 50. 100. 150. 200. 250. 500. 550. 400. TIME (SEC) P n 3-12

I f I j 1 FIGURE 3-3 l DIABLO CANYON TRANSIENT RESPONSE TO STEAM LINE BREAK AT EXIT OF STEAM GENERATOR WITH WATER INJECTION AND OFFSITE POWER .5' 3 25 - M5 d5 .2 [ \\ -gs.15 CC LaJ SW E'.05 0. O. 50. 100. 150. 200. 250. 500. 550. 400. TIME (SEC) 3 D' 3-13

i i ) FIGURE 3-4 DIABLO CANYON' TRANSIENT RESPONSE TO STEAM LINE BREAK AT EXIT OF STEAM GENERATOR WITH WATER INJECTION AND OFFSITE POWER l 1 INITIAL STEAM FLOW IS 9500 LB/SEC FROM THE FAULTED STEAM GENERATOR AND 2900 LB/SEC-(966 PER SG) --D' ~ FRCM THE INTACT STEAM GENERATORS

2. 5 -

5 58 2. gz rs -1.5 -)! Is M5 1. M INTACT STEAM GENERATORS w FAULTED STEAM GENERATOR ONLY b 5 g O. O. 50. 100. 150. 200. 250. 500. 550. 400. TIME (SEC)

• Nominal - 4I32 lbm/sec plant total C

D 3-I4

FIGURE 3-5 DIABLO CANYON TRANSIENT RESPONSE TO. STEAM LINE BREAK AT EXIT OF STEAM GENERATOR-WITH WATER INJECTION AND OFFSITE POWER s LS - mu i_ N 2300 PPM BORON REACHES CORE AT 83 SEC M 05 - O C -o.5 - C v -L5 - ~

• 2 1

O 10 0 200 300 400 TIME (SEC) 9 3-15

+ p v-: ' > .l 3,.. FIGURE 3-6 DIABLO CANYON' g . TRANSIENT > RESPONSE.TO STEAM LINE BREAK AT--EXIT OF STEAM a GENERATOR WITH WATER INJECTION:AND WITHOUT POWER -l 2 L-t is 6 0 0 '. - w C.c x- $ C 500.- s. EU ~ u => 400. 8e vuw . m o. S5 500. ur 4W x 200. 400. D. 50. 100. 150. 200. OED. 500. SED. TIME (SEC) 3-16

L FIGURE'3-7 l DIABLO CANYON io : TRANSIENT RESPONSE TO STEAM LINE BREAK AT EXIT OF STEAM GENERATOR WITH WATER INJECTION AND WITHOUT POWER 2 Ed 2500. PRESSURIZER EMPTIES AT 18 SEC E 2000. - S w W 1500. a. >= 5 1000. c a 8 500.. a: Sy D. g D. 50. 100. 150. 200. 250. 500. 550. 400. TIME (SEC) m O 3-17

i l FIGURE 3-8 DIABLO CANYON TRANSIENT RESPONSE TO STEAM LINE BAEAK AT EXIT OF STEAM GE!44RATOR WITH WATER INJECTION AND WITHOUT POWER ) .b U.25 5* ur .2 LL. O 3 z F-;$s.15 o Wz

• i.

%W ou UE .05 0. D. E0. 100. 150.

200, 250.

500, 550.

4CO. TIME (SEC) b a 3-18

77 - _ _ - g.c i ![ h FIGURE 3-9 .i X4 DIABLO CANYON TRANSIENT RESPONSE TO ST.EAM LINE BREAM AT EXIT OF' STEAM .] 9 GENERATORLWITH WATER. INJECTION'AND WITHOUT POWER i r INITIAL' STEAM FLOW IS 9500 LB/SEC FROM THE FAULTED STEAM GENERATOR AND 2900 LB/SEC (966 PER SG)

1-5.

FROM THE INTACT STEAM GENERATORS U 2.5' (E 32-Se. 2. g.l.5 j*. mw E! INTACT STEAM' GENERATORS wL 5 FAULTED STEAM GENERATOR ONLY f. i a. w e. m D. EC. 100. 150. 200. 250. 500. 550. 400. TIME (SEC). L

• Nominal - 4132 lbm/sec plant total p:

3-19 a

e ~-. FIGURE-3 ' g; ' :DIABLO' CANYON TRANSIENT RESPONSE TO STEAM LINE BREAK AT. EXIT OF STEAM GENERATOR WITH WATER INJECTION'AND WITHOUT POWER, i 4 f ' {. '.' f 'l 1 .I [ :: ] 1 ts _ ~ 4 ' i_ 2300' PPM BORON REACHES CORE AT 98 SEC i <3 J l>Et 05 - w ~O >=- l w

i. o O

'1 aC - 1 W - -1 . g - 'I -r M g. 1 TIME -(SEC) i l' i P. I IJ 3-20 s I \\ L--- j

k FIGURE 3 DIABLO CANYON

• f TRANSIENT RESPONSE;TO STEAM LINE BREAK AT EXIT OF STEAM

~ GENERATOR WITH WATER INJECTION AND OFFSITE POWER 4 3o - as as - 24 - R. CASE B 20 - w ,a-g g_ g W.. ' '10 - ' o 's-v 4-2- i O TIME (SEC) q, 4 ~3-21 4-'

FIGURE 3 ' . DIABLO CANYON E . TRANSIENT RESPONSE TO STEAM..LINE BREAK AT EXIT ~OF STEAM GENERATOR WITH WATER INJECTION AND WITHOUT POWER so - m-m- 7 rr - a m-a- m- ,4 _ n-CASE D z: a-m- - O lo - - 8D e-a. w .a-7- u 5- + 4 s-1- p i i i O 30 0 300 300 6 TIME (SEC) l h 3-22 .1i I. ..i..i. i..i.-:_....... _. 1 -i 1i i.

E

• qi.:.

4-l- l 1 -] FIGURE 3-13 -^- DIABLO CANYON F -. TRANSIENT RESPONSE' FOR A. STEAM-LINE BREAK EQUIVALENT;TO

228 LBS/SEC AT,1015 PSIA WITH OFFSITE POWER A

l 2 Q 2500.-

b

.g: 2000. 8 $.-1500.- n. >==. -{ 1000.- N .a w 8 u-500.- .a

S :-.

u-0. 4 0'. 100. 200. 500. 400. 500. 600. .x TIME '(SEC) D D 3-23

,, J fi,; ' f t-9 '3 I f',,, . FIGURE 3-14 1' DIABLO. CANYON J, TRANSIENT RESPONSE FOR A STEAM LINE BREAK EQUIVALENT.TO ~l =228 LBS/SEC AT:1015 PSIA WITH OFFSITE POWER .r e .n $. G00. - E' _g.500. n y . g-

5. 400.

+ i. .w n'. 500. w ~l -E -200. 8 0. 100. 200. 500. 400. 500. 600. TIME (SEC) 9'.,' F 6 3-24 l

FIGURE 3-15 DIABLO CANYON TRANSIENT RESPONSE FOR A STEAM LINE BREAK EQUIVALENT TO 228 LBS/SEC AT 1015 PSIA WITH OFFSITE POWER 2-L5 - k 2300 PPM BORON REACHES LOOPS AT 328 SEC ib 0.5 - 0 -0.5 - m y r g -i - = .t5 - ' '2 O 200 400 goo TIME (SEC) e o. 3-25 i .c

WESTINGHOUSE PROPRIETARY CLASS 3 4.0 INSIDE CONTAINMENT RESPONSE Introduction The impact'of BIT elimination on the Containment Mass and Energy Release / Pressure and Temperature Response is analyzed. The analyses determine if containment pressure will remain below the design limit of 47 psig. They also determine if containment temperature response will remain below that of the previous analysis which is documented by FSAR figures 6.2-47 and 6.2-49 and 0 by-the, peak temperature value of 339 F in Sect. 3.11.2.1. Analysis The LOFTRAN computer code, (Reference 4), was used to generate the mass and energy release to the containment for a full spectrum of postulated accidents which include large double-ended guillotine breaks at two locations in the steam line, small double-ended breaks in the steam line, and split breaks in the steam line. Each break is analyzed for four plant power levels: 102%, 70%, 30% and 0% power. In addition, a spectrum of single failure criterion was imposed on each~ postulated break scenario. The single failures that are assumed are: 1) failure of the MSIV and check valve in the faulted loop;

2) failure of the feedwater regulating valve in the faulted loop; and 3) failure of auxiliary feedwater pump runout protection to the faulted loop.

Additionally, the full spectrum of breaks / failures is enveloped to the mort limiting of "New' or "Old' steam line break protection logic. The methodology employed is consistent with Reference 8. The mass and energy releases to containment are input to the C0C0 computer code, (Reference 7), which determines the containment pressure and temperature response. Table 4-1 lists. important containment design and operating conditions used in the analysis. 4-1

WESTINGHOUSE PROPRIETARY CLASS 3 Each LOFTRAN mass and energy release is analyzed to envelope two different plant configurations t~ initial conditions. The two configurations are determined by Diablo ( )n Tech Specification 3/4.7.12, (Reference 2). Essentially, two CCW heat exchangers will be in-service whenever ocean water tempei nture exceeds 64*F, and only 1 CCW heat exchanger will be in-service when ocean temperature is less than 64*F. These. configurations impact the cooling capabilities of the containment fan coolers which are cooled by CCW flow. When one of the previously mentioned three failures is assumed, no single failure is assumed as input to C0CO. For LOFTRAN releases where no failure is assumed, the single failure which is input to C0C0 is the failure of one diesel generator. This will impact containment heat removal capability by decreasing the number of operating spray pumps, by decreasing the number of operating fan coolers, and by decreasing the available ultimate heat sink heat transfer capability by decreasing the CCW/ASW systems flow rate. Additionally, each event is analyzed for a coincident loss of offsite power. For the steamline breaks inside containment analysis, the following safety features were taken credit for: reactor trip, steamline ist tion, feedwater isolation and safety injection. These events reduce' or teren ate the mass and energy releases to containment. The signals that provide this protection are:

1. Reactor Trip High 1 Containment Pressure (SI) l Overpower Delta T l

High Differential Pressure (SI) (Old Steam Line Break [0SLB] Protection) High Steam Flow coincident with Low Steam Pressure (SI) (OSLB Protection) Low Steam Pressure (SI) (New Steam Line Break [NSLB] Protection) l 4

2. Feedwater Isolation i'

l Due te Safety Injection Signal l 4-2

o, L WESTINGHOUSE PROPRIETARY CLASS 3 '3. Safety-Injection -c 'High 1 Containment Pressure Low Pressurizer Pressure High Differential Pressure (0SLB Protection) _High Steam Flow coincident with Low Steam Pressure (0SLB Protection) Low Steam Pressure (NSLB Protection)

4. Steamline'Isclation High 2 Containment Pressure High Steam Flow coincident with Low Steam Pressure (OSLB Protection) 1 Low Steam Pressure (NSLB Protection)

The.setpoints which were assumed in the analyses are as follows:

1) High Differential Pressure - 200 psi

2) Low Steam Pressure - 459 psia

3) Low Steam Pressure - 15 psia (for sensitivity cases)

4) Low Pressurizer Pressure - 1695 psia 5).High 1 Containment Pressure - 6.2 psig

6) Overpower Delta T - varies as given below SETPOINT - DLTNOM(K4-K5(TAVGI-TAVNOM)-K6(T3S/(1+T3S))(TAVGI)

DLTNOM - 66.06 TAVNOM = 577.6 l. K4 - 1.1453 K5 - 0.0000 K6 - 0.0000 T3 - 10.000 TAVGI - Indicated Taverage Q'

7) High 2 Containment Pressure - 26.7 psig 4-3

WESTINGHOUSE PROPRIETARY CLASS 3 8);High S' team Flow - varies with power level as given below Power level Setooint w (fract nom power) (fract ' nom flow)- 0.0000: 0.5727 0.2000-0.5727 0.3143' O.7295 0.4286' O.8509 0.5429 0.9537' 0.6572 1.0445 0 715 1.1266 0.8856 1.2023 1.0000 1.2727 1.2000-1.3862 .The BIT and associated piping are conservatively modeled as being in place and full of 0 ppm water. The BIT water volume used is 139 ft3. The delay time. for delivering the 2300 ppm borated water to the primary system covers a 23 -second pure delay. This is the' delay between when.the process parameters m being measured reach the setpoint to the time when the charging pumps are at full flow. In addition to this delay, there is an additional delay to purge the O ppm water between the RWST and the primary system. The modeling of delay times and volumes is consistent with the current Technical Specifications and allows.for the BIT to be removed completely, bypassed, or remain in place, so long as the volume filled does not exceed 139 ft3.and the SI pumped flow rate performance remains the same. l t 1 4-4 l }

WESTINGHOUSE PROPRIETARY CLASS 3 i Results. m The results reported reflect, on a case by case basis, the more limiting CCW 4-heat exchanger configuration and the more limiting steam line break protection logic. The containment pressure transient response.for the limiting case (large break, feedwater regulating valve failure, 30% power) is given as Figure 4-1. There is no single limiting run for FQ temperature purposes over the' duration of the MSLB transient. For this reason the temperature response results are reported in two fashions:

1) Figure 4-2 and Table 4-2 are the aggregate temperature response data that envelopes each and every break scenario.

2) Figures 4-3, 4-4, and 4-5 are the decomposition of this curve into constituent break responses, each with peaks at different points in time. These 3 curves combined together form the enveloping curve. The temperature transient of a given break will be enveloped by the combined curve, but may or may not be enveloped by one of the 3 constituent cases.

"Old" and "New" Protection Logic Q The input parameters used were selected to model either form of protection logic in a conservative fashion. Based on the actuation signals obtained, s either protection system is covered by the current analyses. Low Steam Line Pressure Setpoint Sensitivity As a part of the BIT removal analysis, a study was done to determine if it would be possible to reduce the Low Steam Line Pressure Setpoint. All runs which took credit for the existing LSP setpoint actuation were rerun. It was determined that the setpoint could be substantially reduced with only a minor increase in the temperature envelope, and with the containment pressure limit a still within the design limit. H l 4-5

WESTINGHOUSE PROPRIETARY CLASS 3 The study shows that the.setpoint may be reduced to 15 psia with no impact on the BIT removal MSLB peak pressure. The temperature EQ envelope is slightly worse with one additional case' contributing toward the overall envelope. Figure ~4-6 is the new enveloping temperature curve and it represents the sum of the previous Figures 4-3, 4-4, and 4-5 plus the additional Figure 4-7. The

last column in Table 4-2 represents the changes to the data from the earlier curve..

Note that the 15 psia.setpoint is still reached and continues to provide protection for the large breaks since the indicated steam line pressure is modified by a lead / lag function prior to being compared to the.setpoint.~ Since for large breaks, the rate of pressure drop is fast, the lead /hg anticipates a very low pressure being achieved, and the Low Steam Pressure setpoint is reached. Digital Equipment As a part of the work performed, the impact of future installation of digital C equipment was investigated. Specifically,-the installation of EAGLE 21 hardware was considered. It was determined that steam line break protection logic should not change in relation to the important Non-LOCA parameters. This assumes that the design will constitute a " Form Fit and Function" replacement of current analog hardware. Therefore, the analyses performed should continue to remain valid if EAGLE 21 is implemented. FSAR Revisions FSAR markups are provided reflecting the results of the BIT removal analysis. Markups to section 6.2.1.3.8 are provided to replace and update the original (- FSAR work for Inside Containment. A revision to Sectior. 3.11.2.1 is also provided indicating a new peak temperature limit. No corrections to the 4-6 1

' WESTINGHOUSE PROPRIETARY' CLASS'3' addendum work found in Section 6 App 2B are provided sincejthis work was ~ supplementary to the original. 6.2.1.3.8 information. It is possible for this 4* Laddendum's steamline break information to be deleted'in the next FSAR update,- ~ .since-the addendum's steam line break information:is superceded by the markups

• ?

provided. Conclusion The design pressure limit of 47.psig is met. The temperature transient response is provided to evaluate the equipment response to the Inside steam ~ line break transients. The BIT may be eliminated, bypassed, or ' diluted if the temperature evaluation using Figure 4-2 is satisfactory. The analysis bounds either set of steam line break protection logic and, for peak pressure'only, any Low Steam Line Pressure setpoint greater than 15 psia. With the Low Steam Line Pressure'setpoint as low as 15 psia, the enveloping temperature curve for EQ purposes is Figure 4-6. g-9 im 4-7

' WESTINGHOUSE PROPRIETARY CLASS 3 I i TABLE 4-1 Design and Operating Conditions Assumed c-For BIT Removal MSLB Containment Analysis--

1. Cont. Volume-- 2,550,000 ft3 2'LInitial Pressure-- 16.0 psia,

3. Initial. Cont. Temp.- 120.0F

4. RWST Temp.- '100.0F

~ ^

5. Containment Spray--2600gpm/ pump, Operating based on a 80 sec delay following a Containment pressure actuation setpoint of 26.7 psig (including instrument uncertainties) gL

6. Fan Coolers-- Operate-based on a 48sec delay.following a Containment-pressure actuation'setpoint of 6.2 psig (including instrument uncertainties),

each fan cooler performance as given in FSAR' fig. 6.2-32 0

7. CCW Heat Exchanger UA-- 4,850,000 BTU /hr F calculated from FSAR Table

- 9.2-3

8. Passive Heat Sinks-- Areas, Thicknesses from FSAR Table 6.2b-29.

9. CCW, ASW flows, Ocean Temp-- per table below e

,c 1 4-8 r. . s..

i WESTINGHOUSE PROPRIETARY CLASS 3 Table 4-1 (cont.)

10. Minimum Operating Safeguards Equipment-- per Table below:

Config. 1* Config. 1 Config. 2* Config. 2 w/o Diesel w/ Diesel w/o Diesel w/ Diesel Failure Failure Failure Failure CCW PUMP 3 2 3 2 TOTAL FLOW ** 11000 8000 12000 11000 ASW PUMP 2 1 2 1 TOTAL FLOW ** 10000 9700 19400 9700 CCW HX 1 1 2 2 CS PUMP 2 1 2 1 FAN COOLERS 3 2 3 2 config. 1-Tocean=64F, config. 2-Tocean-70F

• • flow in gpm 4-9

'i. w

WESTINGHOUSE' PROPRIETARY CLASS.3 TABL'E 4-2 1

PGE/ PEG BIT-REMOVAL MSLB ANALYSIS' LIMITING CONTAINMENT. TEMPERATURE VALUES WITH TIME' >J TIME (SEC) , HIGHEST TEMPERATURE HIGHEST TEMPERATURE. WITH CURRENT LSP WITH'STEAMLINE . SETPOINT'(*F) SETPOINT-15 PSIA - 0.1 120 120 1-130 131

• 9i

'191 199

• 33 294 296
• 49 328-

'332

• H57 339

-343

• l 65 344 347
• 73 345J 345 89 336 336 97 332 332 105 327 327

. 113 325 325 121: 322 322 141 315 315 161 308 308 181 307-307 281 322 322 381 312 312' 491 278 278 581 263 263 681 264 264

• changed value from previous column due to. lower setpoint I^

6 4 y-4-10 '2 m.

FIGURE 4-1 DIABLO CANYON BIT REMOVAL MSLB PRESSURE TRANSIENT LIMITING PEAK PRESSURE CASE LARGE DER, FEEDWATER REG. VALVE FAILURE o 30% POWER o 46.55 psig @ 243.7 sec 50. A d _'# ' ,j [ 's [. N 42. T l / \\ l / E5. I e t o j h IC. / 0 P 5. y2 k / 5 / o aE 20. / c E '/ v te. s /' 10. / / / / 5. 7 get je ICE !PI TIME ISEC0!C5 i e 4-11

p _r l' l- -l L. FIGURE 4-2 DIABLO CANYON MSLB EQ TEMPERATURE ENVELOPE a. HIGH TEMP 350.0 l l ll l l Peak Temperature =345 F / 7 300,6 ____d / C \\'- O 250.0 %i 6 200.0 A. x / / 150.0 G / I I 100.0 0.10 1.0 10.0 100.0 1000.0 TIME (SEC) PGE '11T REMOVAL MSLB ANALYSIS LIMITING TEMPERATURES o O 4-12 m.

1 l' l-FIGURE 4-3 DIABLO CANYON BIT REMOVAL MSLB TEMPERATURE TRANSIENT SMALL DER, MSIV FAILURE 102% POWER e 1 1 I i i s ;i a if l j i 0 Peak Temperature =345 F tie. ~ l } / '\\ Etc. f / i see. \\ 262. E

• s

/ \\ A i U e e. s r 8 I E l g22e. E yree. 162. / 1EE. s / Id t. / l 1 j M-12c. 10' 10 10 105 T I P'E I E E CCN:51 e 4-13 l

FIGURE 4-4 I j DIABLO CANYON BIT REMOVAL MSLB TEMPERATURE TRANSIENT i SMALL DER, CST (DIESEL) FAILURE l 70% POWER e .j c see. l b E* / \\ / \\ d ECO. y/ \\! H \\ l 260. \\ u / i Jl 1 I' 2ee. / t' / y ## / l i' / l l ? 222. E = f l' E 2ca. j a / iii

I

/ I.

JE2, g)

I l IE2. / ,A I 3^2. f l - + l 122. g6 30 102 12E T]ME iSECCCS1 l 4-14

FIGURE 4-5 DIABLO CANYON BIT REMOVAL MSLB TEMPERATURE TRANSIENT SPLIT BREAK MSIV FAILURE 102% POWER o c 542. 522. [ / see. / 262. / \\ m 2. \\/ t I u 24c. h E h222. h E 202. W e IE2. / ue. / / / 142. / / / i 3,,, 100 10 102 gg! TIME ISECON05) e - l 4-15

l i FIGURE 4-6 DIABLO CANYON MSLB EQ TEMPERATURE ENVELOPE Q. HIGH TEMP 350.0 gg l l } j j j; j i 41 0 l Peak TeInperature=347 F l \\([..] l lMl;I'- ' k ~--'l lF'"T'! a i i O .300.0 T T I'~ ~ ~~ ~ ~~ { { m l l \\- I i H itt jj N 250.0 "t -- T. " ~~~ T~ 1T ~ m i m j 1 m -l m I i i e j; $ 200.0 ~Ig ~"- -~ 2~~ ~~ ~ ~~ ~~~l~" ~ i m I i / / l [- 150.0 / i i NOTE: CURVE INCLUDES REDUCED h LOW STEAMLINE PRESSURE i SETPOINT(15 PSIA) CASES } t!h ! I j i 100.0 0.10 1.0 ' 10.0 100.0 1000.0 TIME (SEC) PGE BIT REMOVAL MSLB ANALYSIS LIMITING TEMPERATURES a G 4-16

3.,

p 1

P., - ~.3 ~ c ,k'

01. :

FIGURE-4-7 O c DIABLO CANYON O BIT REMOVAL MSLB TEMPERATURE TRANSIENT SMALL DREAK MSIV FAILURE 102% POWER LSP SETPOINT=15 PSIA

a. :

? Ecc. l l 1 1 I I I iili -i l' I i 0 Peak Temperature =347 F .s. 5.:e. s / ) \\ 520. -.si 522.- \\ i m. 4.S \\. ./-- . [ gsc, ~ )/ \\ / \\ Y e. V 3 2 / D 22c.- . i / ! I I f220. / / 160. / / 1ED. / -242. 1 122. je ig2 3g! O'- 100 TIME ISECOt4DS) a. :. 9 4-17 !i._.._._._--._-._--.__._.-.._.-...

WESTINGHOUSE PROPRIETARY CLASS 3 5.0 DOCUMENTATION FOR BIT REMOVAL 1 I FSAR Sections The necessary changes to Chapter 15.0 of the Updated Diablo Canyon FSAR (Reference 1) are provided in Appendix A. The changes provided are the same as those that would be expected to be provided for the Vantage 5 RTSR as they relate to the Steam Line Break transients. Changes for Chapter 6, Section 6.2.1.3.8, dealing with Containment Analysis for Steam Line Rupture are also provide in Appendix A. Noted that the addendum to Chapter 6 which was added describing recent containment calculations for steam line break is now superseded to the extent the addendum relates to steam break by the markups as provided in Appendix A for 6.2.1.3.8 as the original section. Section 3.11.2.1 has also been updated based on the results obtained in this analysis to reflect a new peak containment temperature. All mark ups are relative to Revision 3 of the FSAR. Functional Requirements No changes to the functional requirements were identified as being needed for the BIT removal. Precautions Limitations and Setpoints Document No changes to the PLS document were identified as being needed for the BIT removal. Technical Specifications Appendix B contains marked up copies of the current Diablo Canyon Technical Specifications that reflect the changes necessary to remove the BIT from service. 5-1

WESTINGHOUSE PROPRIETARY CLASS 3 6.0

SUMMARY

AND CONCLUSIONS Plant specific analyses have been performed for the Diablo Canyon Units' ~ steamline break transients and have shown that the Boron Injection Tank may be eliminated provided that analysis demonstrates acceptable equipment qualification with the data provided. The acceptability of the change is made possible in part by the change in the Westinghouse internal criterion for the Condition II steamline breaks. It remains for future analysis to demonstrate that the temperature curves given in Section 4.0 for the Inside Containment Analysis are acceptable. Future analysis should also demonstrate that the essential equipment outside containment is qualified for the appropriate transients as presented in Appendix A of Volume 2, in terms of mass and energy release. Once this is accomplished then the BIT may be removed from service after the Technical Specification changes have been approved. The analyses performed allow for the reduction of the Low Steam Line Pressure Setpoint to a value supported in the analyses of 15 psia. The final plant setpoint must consider instrument uncertainty for this analysis value. The additional temperature data in Section 4.0 and the additional cases provided in I Volume 2 are provided to support the Equipment Qualification effort in this regard. F The analyses have been performed with the intent of their continuing to be applicable for the future transition to Vantage 5 fuel. In addition, it is expected that the analyses performed will remain applicable if the digital Eagle 21 protection system is installed. Because of this early effort, final validation effort upon implementation of the changes is expected to be minimal. 6-I I

WESTINGHOUSE PROPRIETARY CLASS 3

7.0 REFERENCES

i 1. Diablo Canyon Updated Final Safety Analysis Report, USNRC Docket Number 50-275. 2. Diablo Canyon Technical Specifications. 3. FSAR Ammendments anel Technical Specifications for Beaver Valley, Byron /Braidwood, Catawba, J. M. Farley, McGuire, Callaway, and Turkey l. Point, for BIT removal 4. WCAP-7907, T.W.T. Surnett, et. al., "LOFTRAN Code Description," October, 1972. 5. ANS1/ANS-5.1-1979, " Decay Heat Power In Light Water Reactors", August 29, 1979. 6. WCAP-9226-R1, " Reactor Core Response To Excessive Secondary Releases", S. D. Hollingsworth and D. C. Wood, January 1978. 7. WCAP-8327, " Containment Pressure Analysis Code (C0CO), Bordelon, F. M. and E. T. Murphy, July 1974. 8. WCAP-8822-SI-P-A, " Mass and Energy Releases Following a Steam Line Rupture", R. E. Land, September, 1986. W 7-1

WESTINGHOUSE PROPRIETARY CLASS 3 APPENDIX A FSAR REVISIONS FOR BIT ELIMINATION 4 4 = -F A-1

WESTINGHOUSE PROPRIETARY CLASS 3-1 I APPENDIX A FSAR CHAPTER 3 REVISIONS O e S A-2

DCPP UNITS 1 & 2 ISAR UPDATE Trescribed limits during the test, the necessary corrections will be made or minor maintenance perf ormed and the unit retested immediately. Satisf actory perf ormance of the remaining redundant component (s) is proof of the availability of that safety f eature, and it is not necessary to adjust plant lead during the brief period that a saf ety f eature component way be out of service, e The environmental conditions to which ESF equipment may be exposed, either during normal -) operation or following an accident, are given below: 1 3.11.2.1 Environmental Conditions inside the Containment Norpel operating pressure, dry bulb temperature, and relative humidity inside the containment will be 14.7 psia, a maximum of 120'F, and 53%, respectively. j In the event of a LOCA, the pressure, dry bulb temperato1, and humidity inside the containment can approach containment design pressure 2nd tocerature, as discussed in Section 6.2.1, Pressure and temperature profiles are given in Appendix 6.2B of the FSAR Upate. 1 For a postulated PSLB accident, pressure and temperature profiles are shown in Appendix 6.2B of the F5aR Update. The humidity condition is 100% under the MSLB event. Test conditions enveloping these values were used to Qualify safety-related equipment. The ( analyses used to develop the service envelopes f or environt.wrie.1 qualification cie-4y with l the requirements of NUREG 05BB. The peak pressure and temperature conditions taken from these service envelopes are as follows: 4 Mar. Temp. 'F Man. Press, osic Humidity. % LOCA 272 46.91 100 MSLB p{ g 100 Chemical spray has been included in testing programs for eauipment located inside the containment. The expected chemical composition and pH condition in the LOCA are described in Section 6.2.1. Containment flooding following a LOCA would have no adverse effect en safety-related electrical control and instrumentation circuits and devices. An evaluation supporting this conclusion is included in Appendiz 3.11A. l-i e 3.11-3 September 1987 Revision 3 A-3

i l-l; WESTINGHOUSE PROPRIETARY CLASS 3 l l l APPENDIX A FSAR CHAPTER 6 REVISIONS 1 l l 6 9 o O A-4

DCPP UNITS-1 & 2 FSAR UFCAlE (4) From curve lA on figure 6.2-8 (assuming RHR recirculation begins at approximately 1g00 seconds when 105,000 gallons remains in the RWST), the maximum containment recirculation sump water temperature at the start of pp post-LOCA RHR recirculation would be 24B'F. 6.?.1.3.8 Containment anelvsis - Steam line Ruoture t Coltainmentpressureresponse'toaspectrumofsteamlineruptureswasevaluatedI N, assuming they are coincident with a single f ailure of either;hne en rgency dieselgeryweho en at pr a u n o jeavy r p sf,C 9 2)emainsteamlineisolationvalve,Q# a feedline control valveg F r each singfe f ailure the .- W."'t most limiting initial conditions were assumed, thereby ensuring that the most severe case is analyzed. TWhe(4 h Should a steam line rupture inside containment, the steam generator closest to the break will lose its fluid inventory. Rapid depressurization of the steam generator will cause a sharp increase in the feedwater flow to the steam generator. This flow will be maintained until the feedwater pumps are tripped and the feed line control valve is closed. If the feed line control valve is postulated to fail, flow will continue to enter the steam generator until the back-up feed isolation valve is closed. This flow may be accompanied by flashing of the high-energy fluid in the feed lines. Reverse flow from the other steam generators is prevented by a check valve located in each steam line outside the containment. Should this valve be postulated to fail, the other steam generators can blow down until the steam line isolation valves are closed. The following systems are available to ensure that the reactor is tripped and the steam and feed lines are isolated from the break. a (1) Reactor trip on any of the following: (a) High neutron flux (b) High flux rate (c) Overtemperature AT (d) Overpower AT (e) Low pressurizer pressure (f) Low-low steam generator water level (g) Safety injection signal ( o 6.2-la September 1986 Revision 2 A-5

DCPP UNITS 1 & 2 FSAR UPDATE (h). Low steam generator level and steass-feed mismatch. (2) Safety injection system (SIS) actuation (and reacter trip) on any of the following signals: (a) Two-out-of-three high containment pressure (b) Two-out-of-four low pressurizer pressure (c) High steam line flow in two maio stama lines (one-out-of-two per line) in coincidence with either low-low RCS average temperature or low steam line pressure in any two lines 1 (d) High differential pressure between steam lines. (Any one steam line lower than aqy two of the other three steam lines.) (3) 5 team line isolation valve actuation: (a) High steam flow in two main steam itnes coincident with either low steam line pressure in any tuo lines or low-low primary coolant average temperature (b) Two-out-of-four high-high containment pressure. a (4) Feedwater system isolation (inc1sding amergency closures of the feedwater regulating valves, closure of feedwater t'ypass valves, trip of the feedwater 1 naps, arid closarre of the feedwater isolation valves) on: (a) Ary safety injection signal (b) Two-out-of-three high-high steam generator level in any one steam generator (c) Reactor trip erith low T,,9 (FW evpulating valves only). l A series of steam line brusts eere analyzed to determine the most severe breat condition for containment temperstare and pressure response. The following assumptions were used in the analysis: (1) Break sizes erete assumed to be eitner double-ended ruptures or split ruptures. Ilouble-ended irreats were assumed to occur at the outlet of one steam generator swd downstream of the flow restrictor. l 6.2-15 September 1985 Revision 1

DCPP UN!TS 1 8 2 FSAR UPDATE v S& fS g:%cr c e f r c i s. eA 5 4 ro.<. (2) Blowdown from the broken steam line is assumed to be satwatedateem. r_ (3) The reactor will trip on high steam line differential pressure, or SIS e actuation (with resultant reactor trip and feed line isolation); steam line isolation will occur on high steam line flow coincident with low steam line A high containment pressure signal will also be actuated to cause pressure. steam line isolation. The main feed control valves and the steam line isolation valves were assumed to have a 5-second closure time. The backup feed isolation valve (for the failed feed cortrol valve case) has a makfiaum 60-second closure time. No credit was taken for reduced flow through the valve during closure. (4) plant power levels of 102% of nominal full load power, 705 of nominal full load power, 30% of nominal full load power, and zero power were considered. Power level is based on Unit 2 power. (5) The double-ended ruptures were evaluated for,fulg' double-egdef guillotine 'j '7f .,,o,e ruptures, 3.69 and 1.4 square feet [heYpYit ruptNe's$re ev#afu'ated$t /cef ~ 0.96, 0.942, 0.908, and 0.86 square feet. Table 6.2-49 identifies the limiting cases for each break size and power level analyzed. Steam line flow restrictions in the steam lines limit the effective break area of a full double-ended pipe rupture to a maximum of 1.4 square feet per steam generator if the break occurs downstream of the restrictors. Upstream, the outlet nozzles of the steam generator limit the effective break area to 3.69 square feet. O (6) Failures of a main steam isolation valve, a feed line control valve, and a diesel generator and auxiliary feedwater runout are considered. (7) The auxiliary feedwater system is manually realigned by the operator after 10 minutes. h*" "

• M 'l l

(D fu Lv /s.Ne-cvAr*/'~/ % es,Ll'rO/'*/h ' > -2 % Main steam line break flow modeling of the blowdown of the steam generators and plant steam piping was as follows: (1) Steam Generator Blowdown Break flows and ent ies from the steam generators were calculated using the Westinghouse : p code. Blowdown mass and energy release were bN determined using the b code, including effects of core power generation, ~ a main and auxiliary feedwater additions, ESFs, RCS thick metal heat storage, and reverse steam generator heat transfer. 6.2-16 l A-7

DCPr UNITS 1 & 2 F5AR UFC#TE (2) Steam Plant Piping Blowdown The contribution to the mass and entrgy releases from the secondary piset steam piping was included in the mass and energy release calculations. For all ruptures, the steam piping volume blowdown begins at the time of the break ared continues at a unif orm rate untti the entire piping inventory is released. The flowrote was determined using the Moody correlation, the pipe cross-sectional area, and the initial steam pressure. Following the piping blowdown, reverse flow from the intact steam generators continues to simulate the reverse steam generator flow prior to steam line isolation. The following single failures were evaluated to determine the limiting set of conditions for this analysis: (1) failure of a main steam line isolation valve increases the volume of steam piping which is not isolated from the break. When all valves operate, the piping volume capable of blowing down is located between the steam generator and the first isolation valve. If this valve fatis, the volume between the break and the isolation valves in the other steam lines, including safety and relief valve headers and other connecting lines, will feed the break. (2) Failure of a diesel generator would result in the loss of one containment safeguards train resulting in minimum heat removal capanility. (3) Loss of the auxiliary feedwater runout protection would result in a more conservative auxiliary feedwater flowrate and, consequently, a greater steam release. I (4) Failure of a feedwater control valve e rM riy resultsin additional inventory in the feedwater line which would not be isolated from the steam generator, a The mass in this volume can flgsh into the steam generator and exit through Sb d is aho c orst.itIUed 9"ds/ the break. An anuensed hebakr'lscMen A met (n a dr..d.t im d hhne e( ut. (-ce E ak e t oebel v6bt. li set #i

h...W.3 as e9 5 %..,.,,_ __,

pf4 _ a.s ,...,,i..pp. m .n - M: p;;. in in ispM o.e-> > anc b.z-as. To analyze the containment response to the rupture of the main steam line, the pressure, temperature, and humidity of the containment atmosphere prior to the postulated accident, as well as values for the temperature of the service water and RW51 solution, are assumed, along with the initial water inventory of the RWST. All of these values are chosen conservatively, as shown in Table 6.2-50. I l s 6.2 11 A-8 _ _A

DCPP UNITS 1 & 2 FSAR UPDATE r

• 4,,94 Le, ;

} i s ,eg g ],c o r in eachlcf the transients, the safeguards systerms shown in Table 6.2 50 are assumed to operat W indicated. Tables 6.2 53 and 6.2-54 present the accident chronology for the limiting steam line ruptures. The assumed spray flowrate is based on M twc trains operating. o As in a LOCA event, the significant heat removal sink during the early portion of the transient is structural heet remeval. The methodology used para 11e's that described in Section 6.2.1.3.2 above. Provision is made in the containment pressure transient analysis for heat transfer through, and heat storage in, both interior and exterior walls. Every wall is divided into a 1&rge number of nodes. For each node, an energy conservation equation, expressed in finite difference form, accounts for transient conduction inte and out of the node and temperature rise of the node. Also shown in Table 6.2-50 is the summary of the containment structural heat sinks used in the analysis. 1ht heat transferred from the containment atmosphere to the containment structure is calculated following primarily the work of Tagami. The value of the heat transfer coefficient between the containment atmosphere and the containment structure increases parabolica11y to peak at the time of steam line isolation. The value of the heat transfer coef ficient then decreases exponentially to a stagnant value which is a function of steam-to-air weight ratic. As with the LOCA analysis, containment spray plays a significant role. During the injection phase of operation, the containment spray pumps draw water f rom the RW5T and sprays it into the containment through nczzles mounted high above the operating deck. As the spray droplets f all, they absorb heat f rom the containment atmcsphere. Since the water comes from the RWST, the entire heat capacity of the spray free the RW5T temperature to the temperature of the containment atmosphere is available for energy absorption. L as Wi. iw6ir6uisiion pnase si I wt ace W t ;,....i6, hi b e. '; a P: ; g d m y;d ' L Wi. 6vnteu - ni m.aero-hil C VC wf.T cwca G t est 4Cfmisc</ca$ prie*' is tecirtuI" W S w,' + t k ov V When a spray drop enters the hot, saturated, steam-str containment environment following a LOCA, the vapor pressure of the water at its surfs:e is much less than the partial pres *sure l cf the steam in the atmesphere. Hence, there will be diffusion of steam to the drop surface 1 l and condensation on the drop. This mass flow will carry energy to the drop. Simultaneously, the temperature difference between the atmosphere and the drop will cause the drop temperature and vapor pressure to rise. The vapor pressure of the drop will eventually l become equal to the partial pressure of the steam and condensation will cease. The temperature of the drop essentially equals the temperature of the steam-sir sixth a. The RCFCs are a final means of heat removal. The main aspects of a fan cooler from the heat removal stardpoir.t are the fan and the banks of cooling coils. The fans draw the dense atmosphere through banks of finned cooling coils and six the cooled steam-air sisture with '~ 6.2 18 A-9 l c i

DCPF UN1151 & 2 f54R UPDAll the rest of the containment' atmosphere. The coils are kept at a low temperature by a constant flow of cooling water. Since this system does not use eats f rom the RW5T, the mode of operation remains the same both before and af ter the spray system and ECCS change to the ^* recirculation mode. . Containment pressure and temperature transients are calculated with the C0CO computer code, sults of the pressure transient analysis of the containment for the secondary si breaks are ented in Table 6.2-49. The pressure and temperature curves fo e most limiting steam brea are presented on Figures 6.2-47, 6.2-48, 6.2 4 d 6.2 50. Tables 6.2 51 and 6.2-52 present mass and energy released duri ese transients. As shown, containment peak pressures and t atures remain w design values. The most limiting MSLB was reanalyr ing credit itional heat sinks. The initial conditions and assumed cent nt safety features were revise the results of this analysis, shown in ndia 6.2B, it was concluded that the R$tB pressure an erature curves in ures 6.2-41 to 50 are more limiting with respect to both containment pres ) \\ ank temperature than the reanalyzed MSLB case. 6.2.1.3 9 Containment Pressure differential During the early stages of a large area LOCA, pressure differentials may be briefly established in the containment. While the geometry of the containment, except for the net f ree volume, has no direct ef fect upon the containment peak pressure, indirect considerations such as the design of structural supports of [5F equipment and the prevention of missile generation make it desirable to calculate the dif ferential pressure transients caused by different breaks. Four cases are of interest: (a) a rupture of an RCS hot leg at the biological shield that results in the maximum dif f erential pressure across the loop compartment walls, (b) a rupture of an RCS hot leg at the reactor vessel norrie weld that results in the marteum reactor cavity differential pressure, (c) a pressurizer spray line rupture that resuits in the maximum pressurizer enclosure differential pressure, and (d) a hot leg break in one of the steam generator loop compartments that yields the maximum pressure dif ferential across the steam generator. These four cases were analyzed using the TRD computer code with an unaugmented homogeneous critical mass flowrate correlation. As a result of comparisons at low pressures between measured critical mass flowrates and predictions using the homogeneous critical flow model, an equation has been developed that conservatively bounds experimental critical mass flowrates by applying an augmentation f actor of (1.2-0.2K) to the homogeneous model flowrates, where K is steam cuality in the upstream compartment. Since critical mass ~' flowrates obtained using the augmentation factor are conservative with respect to experimental data (calculated flowrates are lower than observed), peak compartment pressures calculated using the augmentation f actor would be expected to be conservative (higher than A-10 6.2 n September 1986 Revision 2

INSERT A- 'The results of the BIT-removal MSLB containment reanalysis are sumrarized .in Table 6.2-49. The table includes only the limiting failure case for sach. postulated break size and power event. The overall limiting pressure ~ transient of record is the large break from 30% power coincident with a- 'feedwater regulating valve failure, which results in a peak. pressure of-46.55 psig..The accident with the highest predicted peak temperature of-l ' record is the small break from 102% power coincident with a MSIV failure. The peak temperature for this case is 345 F at 73 sec..The peak ) pressure'and~ peak temperature case curves are presented in Figures 6.2-47 L and 6.2-48.. The accident sequence history for these two cases are presented in Tables 6.2-53 and 6.2-54. INSERT B The Main Steam Line Break accident has been reanalyzed in conjunction with BIT removal. The containment pressure and tg/mperature results are more limiting than previous analyses. INSERT C The analysis bounds both' loss of offsite power and no loss of offsite ~ power as a coincident event.

• h.

A-11

DCPP UNIVS 1 & 2 fSAR UPDA7E 28. H. E. Zittel, ar.d T. H. Row, ' Radiation and Thermal stability of Spray solutions', huel. Technol. 10, 1971, pp. 436-443. ~ 29. .A. C. Allen, The Radiation Chemistry of Water and Aoveous Solutiens, Princeton, N. J., Van Nostrand, 1961.

30. Limits of Flaw.atility of Gases and vapors, Bureau of Mines Bulletin 503, U.S.

Government Printing Office, 1952. 31.

2. M. Shapiro and T. R. Moffette, Hydronen Flammability Data and Aeolication to PWR Loss-of-Coolant Accident. WAPD-SC-545, September 1957.

32. M. G. Zabetakis, Research on the Combustion and Explosion Harards of Hydrocen-Water Vaeor Air Mirtures, AECU-3327, September 1956. 33. R. M. Kemper. Iodine Removal by So av in the Salem Station Containment. WCAP-7952, August 1972. 34. F. M. Bordelon, and E. T. Murphy, tentainment Pressure Analysis Code (C0CO), WCAP-8326, i June 1974. l 35. F. M. Bordelen, SATAh-VI Procram Comprehensive Space-Time Dependent Analysis of Less-of-Coolant, WCAP-8306, June 1974. 36. G. A. Israelson, J. R. van Searen W. C. Boettinger, Reactor Containment Fan tooler Cooline Test Coil, WCAP-7336-L,1969. r 37' Burnett, T. W. T., et al., 'LOFTRAN Code Description," WCAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984. L

38. Diable Canyon Power Plant - Unit 1 Inservice inspection and Testino Procram - The First 10-Year Inspection Interval, September 28, 1981.

l l 39. C. W. Fields Fan rooler Motor Unit Tests, WCAP-7829, April 1972. l 40. U.S. NRC Standard Review Plan 6.2.5, Combustible Gas Control. l i

41. Branch Technical Position CSB 6-2, control of Combustible Gas Concentrations in Containment Followino a Loss-of-Coolant Accident.

i

42. Whyte D.D., Surchell, R. C. Corrosion Study for Determining Hydrocen Generatino From Aluminue and 2ir: Durino Post-Accident Conditions, WCAP-8776, April 1976.

6.2-81 j A-12 l a

DCPP UNIVS 1 & 2 FSAR UPDATE l

43. Westinghouse Project Letter.- 4485, Deanalysis of the Double-ended Pume Suetion Break and Sensitivity to Containment Stray Delav Time. November 15, 1981.

44. (Wst4*thevn h ebit i.itter- [I7 j" # 5'::- t*ne B ^9 An h M M L--.. i) 4 l' //5 ]%

V* /.1~ 6.t,- 2L.%%'A ;as) f.wp fta C.,4 cst cov/cs' an% a' / 4cc ' A r.* . m i5 t /. / 45. F. M. Bordelon, Analysis of the Transient Flow Distribution Durine Blowdown f TMD Codel ~' WCAP-7548, 1970. 46. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2. Appendix A to License hos. DPR-80 and DPR-82, as amended to the date of the most eecent FIAR Update Revision. 47. IEEE-Std-334, Guide for Tree Tests of Class 1 motors Installed inside the Containment of Nuclear Power Generatir.c Stations, 1971. 48. AIC Saf ety Guide !!, Instrument Lines Penetrating Primary Reactor Containment, March 10, 1971. j

49. Westinghouse Letter PGE-6740. Hydrocen Analysis with Additional Aluminum Inventory, May 22,1986.

hh hkWfY ( l f e D( { (Y hY e )o A Q ser (ode (C o c o) ", vvch P q321, si

1979, g

S 1. 6.2-82 5 ' '"" " 8 6 " " ' ' ' "# A-13

OCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-49 CONT AINMEN1 PEAK PRESSURE AND TEMPERATURE FROM STEAM LINE BREAK ANALYSIS l Peak Peak Pressure, Temperature. Steam Line Break psic 'F 1.4 ft2 DER (feedwater i elation valve failure) 102% Power 41.2 262. 3.69 ft? DER (main steam line isolation valve failure) 102% Power 35.2 272. 1.4 f t2 DER (main steam line isol tion valve failure) 70% Power 38.2 288. 1 3.69 ft2 DER (main steam line isolat on valve failure) l 70% Power 35.6 273. 1.4 ft2 DER (main steam line isolation y lve failure) 30% Power 38.3 294. 3.69 ft2 DER (main steam line isolation valv failure) ~ 30% Power 38.4 279.6 1.4 ft2 DER (main steam line isolation valve f aSQure) 0% Power \\ 36.6 296. ~ DER (main steam line isolation valve f ailu\\s) 3.69 ft2 0% Power 39.1 283. 0.86 ft2 split rupture (containment safeguard train f ailure) 102% power 44.4 337. 1 0.908 ft2 split rupture (containment safeguard train failure) 70% power 43.4 339. 0.942 ft2 split rupture (containment safeguard train failure) 30% power 40.4 338. 0.96 ft2 split rupture (containment safeguard train failure) 0% power 35.2 335. \\ A1' We W g t(h s N\\ A-14 \\s 'N i A

n TABLE 6.2-49 CONTAINMENT PEAK PRESSURE AND TEMPERATURE FOR BIT REMOVAL MSLB ' ANALYSIS BREAK SIZE POWER PEAK PRESSURE PEAK TEMP.' / FAILURE / FAILURE 2 ft psig F Large 102 45.3/FRV 263/FRV Large 70 46.47/FRV 266/FRV Large 13 0 46.55/TRV 268/FRV Large 00 44.68/TRV 264/FRV Medium 102 36.18/MSIV 254/MSIV Medium 70' 36.3/MSIV 253/MSIV Medium 30 .40.6/MSIV 281/MSIV Medium 00 37.5/MSIV 254/MSIV Small 102 42.38/MSIV 345/MSIV Small 70 43.1/MSIV 334/MSIV Small 30 41.6 MSIV 318/MSIV f Small 00 30.6/MSIV 248/MSIV .86 split 102-43.56/MSIV 326/MSIV .908 split 70 43.74/MSIV 327/MSIV .944 split 30 46.41/MSIV 328/MSIV .40 split 00 39.00/MSIV 260/MSIV 2 Large=3.69 ft do ble ended rupture Medium =1.4 ft do ble ended rupture Small=.7 or.6 ft double ended rupture-102% =.6 or.5 ft double ended rupture-70% =.4 or.3 ft double ended rupture-30% =.2 or.1 ft double ended rupture-0% FRV=Feedwater Regulating Valve MSIV= Main Steam Isolation Valve A-15

I H DCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-50 Sheet'1 of 4 IN11]AL EDNDITIONS FOR STEAM LINE BREAK ANALYSIS Containment Design Parameters Containment Design Fressure, psig 41 Containment Volume, ft 2,550,000 Initial Containment Pressure, psia M I b-Initial Containment Temperature, *F 120 Refueling Water Storage Tank Temperature (Max), *F 100 Outside Containment Temperature (Max), *r 120 Debembic Refueling Water Storage Tank Inventory, gal 6 3 OI2N Service Water Temperature, 'T 70,G'l % Containment Safety Features Spray System i Number of Spray Trains 2 Number of Spray Trains Operating in 1 Minimum Safeguards Analysis (b>nci Fahe) Number of Spray Trains Operating in 2 Maximum Saf eguards Analysis (otker Failure) Spray Flowrate per Spray Train 26D0 gpm W sx is rer A e yo*p W ' c /', ces ,g cca M w W l p O A-16

f DCPP UNITS 1 & 2 FSAR UPDATE 1ABLE 6.2-50 Sheet 2 of 4 Fan Coolers Number of Fan Coolers 5 Number of Fan Coolers Operating in Min Safeguards Analysis 7. (Otesef Fe:tes) Number of Fan Coolers Operating in Max Safeguards Analysis A' 3 ( e ke Fabre) Initiation Times /Setooint W 0elay After Setpoint Containment Setpoint is Reached, Systee Used see ~ Spray 26.7 psig 71.100 7 0 ' fan Coolers 6.2 psig 48.0 t 1 uclsJ es L ot) *{ O WS* h * ?c"c DkCICI M 'I'!N'Y '*'~E Y A-17

s. OCPP UNITS 1 & 2 FSAR UPDATE 1 ABLL 6.2-50 Sheet 3 of 4 m Passive Heat \\ Sinks O Thermal Volumetric Item # Area, ayer Composition Thickness Conduc., Heat Capac., ft2 Stu _ 8tu 3 hr-ft *F ft ,,7 1 92100 1 Paint 7.5 mils 0.2083 36.86 2 Steel 3/8 in. 28.0 58.5 3 Concrete 1.0 ft 1.04 23.4 2 9700 1 int 7.5 mils 0.2083 36.86 2 St 1 1/32 in. 28.0 58.5 3 '} 9000 1 Paint 7.5 mils 0.2083 36.86 I 2 Steel 1/16 in. 28.0 58.5 4* 18000 1 Paint 7.5 mils 0.2083 36.86 2 Steel 3/32 in. 28.0 $8.5 5 17100 1 Paint 7.5 mils 0.2083 36.86 2 Steel 1/ in. 28.0 58.5 6 50500 1 Paint 7.5 mi 5 0.2083 36.85 3/16 i\\. 28.0 58.5 2 Steel 7 9500 1 Faint 7.5 mils 0.2083 36.85 2 Steel 1/4 in. 28.0 58.5 8 37000 1 Paint 7.5 ratis 0.2083 36.85 2 Steel 3/8 in. 20 58.5 9 -9500 1 Paint 7.5 mils 0. 83 36.85 2 Steel 7/16 in. 28.0 58.5 10 26200 1 Paint 7.5 mils 0.2083 36.85 2 Steel 1/2 in. 28.0 58.5 11 21000 1 Paint 1.5 mils 0.2083 36.85 2 Steel 3/4 in. 28.0 5.5 12 17862 1 Paint 7.5 mils 0.2083

36. 5 2

Steel .20 in. 28.0 58.5 A ~' ( ypt' A-18 \\

OCPP UNITS 1 & 2 FSAR UPDATE TABLE 6.2-50 Sheet 4 g 4 Passive Keat Sinks (Cont.) Thermal Volureetric Ar Layer Composition Thickness Conduc., Heat Capac., Item #

ftga, Btu

_ 8tu hr-ft 'F ft

• F 13 10460 1

Paint 1.5 mils 0.2083 36.85 2 Steel .15 in. 28.0 58.5 14 4325 1 aint 1.5 mils 0.2083 36.86 2 5ttet 0.2 in. 28.0 58.5 Patb 7.5 mils 0.2083 36.B6 15 2300 1 2 Steel 432 in. 28.0 58.5 16 2800 1 Paint 7.5 mils 0.2083 36.86 ? Steel S-3/4 in. 28.0 58.5 17 15525 ") 1 Faint 7.Shils 0.2083 36.86 I 1.0ft\\ 1.04 23.4 2 Concrete 18 12610 1

5. Steel 0.438 in.' s S.6 58.5 19 31400 1

Paint 7.5 mils 0.2083 36.Bb 2 Concrete 1.0 ft 1.04 23.4 20 4000 1 S. Steel 3/16 in. 8. 58.5 (a) In contact with water. ,.4 i i l e v il Ti l i li N ligy ty i :to 2 h ! 11 ili pi ,j ll "U ^ '" E 100 g i i _ i i in j 6 i s ils Nit! it i t - tin w e 6 I t ili n i i i 16 il N 61 i i iih li\\ li I I e iil I h i iiIlit l l t ill I i i i ii ,i i 4, N ii i ii., w i i ieino i i isinn iin i i il N ii. o ini i i, ipp i.n i ii i ii! i i.n I I i lit h t ill i il \\ 1114ll l l I t ;l e !p 66 n i n 1 liti .iii 6 i i ll I illI i Il \\ lli I t ti! 6 i r i t i si , i i i, ii i, s..ii l 10 1 100 101 102 1o3 104 TIME (SECONDS) FSAR UPDATE 1lM UNITS 1 AND 2 QN DIABLO CANYON SITE c FIGURE 6.2-49 CONTAINMENT TEMPERATU E HISTORY FOR 0.908 FT2 SPLI STEAM LINE BREAK-(70% POWER - CONTAINMENT SAFEGUARD TRAIN FAILURE) A-30

50 i s A 1 ao l' l ai i. i s ,I. i t K f 1* m s -( }i t. t/ Ii i ~ i i x it if ,t. W g \\ }i J 4 i E 4 ti f I., D i !i i,,r i.6 ii I! I '?i i t i -- w i., 4 . n i. ,i,,.. u. i ..o. g i. ,e

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i 16 !.i' i t6, f x i 4 .,o, i i > >n s 4 1 o 10 1 100 101 102 103 104 TIME (SECONDS) FSAR OPDATE R UNITS 1 AND 2 W DIABLO CANYDN SITE (i ye ( FIGURE 6.2-50 CONTAINME NT PR ESSUR E STORY FOR 0.908 Ft2 SPLIT STE AMilNE BREAK (70% POWER - CONTAINMENT SAF EGUARD T IN FAILURE) A-31

L WESTINGHOUSE PROPRIETARY CLASS 3-l l 't APPENDIX A FSAR CHAPTER 15 REVISIONS 98 N A-32

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DCPP UNITS 1 & 2 TSAR UPDA1E .In calculating the DNBR the following conservative assumptions are made: 1 (1) Initial conditions of maximum core power and reactor coolant temperatures and I minimum reactor coolant pressure exist, resulting in the minimum initial i margin to DNB (see Section 15.11 l (2) A +5 pcm/*F moderator coef ficient of teactivity for BOL operation in order to provide a conservatively low amount of negative reactivity feedback due to changes in moderator temperature. The spatial effect of void due to local or subcooled boiling is not considered in the analysis with respect to reactivity feedback or core power shape. (3) A high (absolute value) Doppler coef ficient of reactivity such that the resultant amount of positive feedback is conservatively high in order to retard any power decrease due to moderator reactivity feedback. It should also be noted that in the analysis power peaking f actors are kept constant at the design' values while, in fact, the core feedback effects would result in considerable flattening of the power distribution. This would significantly increase the calculated DNBR; however, no credit is taken for this effect. 15.P.12.3 Results l~ figure 15.2-6B illustrates the flux transient following the RCS depressurization accident. The flux increases until the time reactor trip occurs on overtemperature AT, thus resulting in a rapid decrease in the nuclear flux. The time of reactor trip is shown in Table 15.2-1. The pressure decay transiert following the accident is given on Figure 15.2-69. The resulting DNBR never goes below 1.30 as shown on Figure 15.2-10. Average temperature is ~ shown in Figure 15.2-11. 15.2.12.4 Conclusions The pressurizer low pressure and the overtemperature 6T reactor protection system signals provide adequate protection against this accident, and the minimum DNBR remains in excess of 1.30. l l 15.2.13 Accidental Deoressuriration of the main Steam System 15.2.13.1 Identification of Causes and Accident Description The most severe core conditions resulting from an accidental depressuriration of the main steam system are associated with an inadvertent opening of a single steam dump, relief, or safety valve. The analyses, assuming a rupture of a main steam pipe, are discussed in Section 15.4 l l 15.2-34 September 1987 Revision 3

l'. DCPP UN!15 1 & 2 FSAR UPDATE The steam released as a consequence of this accident results in an initial increase in steam - I.. flow that decreases during the accident as the steam pressure falls. The energy removal from the RCS causes a reduction of coolant temperature and pressure. In the presence of a l negative moderator temperature coef ficient, the cooldown results in a reduction of core shutdown margin. ,. s v y 8 b Theanalysisisperformedtodemonstratethatthefollowingcr*terionis[atisfied: Assuming

h. '.

s e a stuck RCCA and a single f ailure in the engineered safety features ([5F)f n.-i" 6... U s -i-- '- ~*::'f'af ter reactor trip f or a steam release equivalent to the spurious opening, with failure to close, of the largest of any single steam dump, relief., or safety valve. The following systems provide the necessary mitigation of an accidental depressurization of the main steam system. (1) Safety injection system (515) actuation from any of the following: (a) Two-out-of-four low pressurizer pressure signals (b) High differential pressure signals between steam lines. (2) The overpower reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safety injection sigr.a1 (3) Redur. dant isolation of the main feedwater lines: Sustained high feedwater flow would cause additional cooldown. Therefore, a safety injection signal will rapidly close all feedwater control valves, trip the main feedwater ) pumps, and close the backup f eedwater isolation valves. r Li'Y 15.2.13.2 Analysis of Effecti and Consequences l 1he following analyses of a secondary system steam release are performed with the'MA%d code: A (1) A full plant simulation to determine RCS temperature and pressure during g)0 g (2) An analysis to ascertain that the reactor does not ^r- ^ : - 't l The followina conditions are assumed to exist at the time of a secondary system break accident. o 15.2 35 A-36

- DCPP Uh1151 & 2 FSAR UPDATE (1) EOL shutdown margin at no-load, equilibrium menon conditions, and with the most reactive assembly stuck in its fully withdrawn position. Operation of ~ RCCA banks during core burnup is restricted in such a way that addition of positive reactivity in a serendary system break accident will not lead to a more adverse condition than the case analyzed. y (2) A negative moderator coefficient corresponding to the EDL rodded core with the most reactive RCCA in the fully withdrawn position. The variation of the coefficientwithtemj te and pressure is included. The k versus gg temperature curve a psi corresponding to the negative moderator y ' { temperature coef ficient 4a-ooed plus the Doppler temperature ef f ect is shown on Figure 15.2-41. (3) Minimum capability for injection of high concentration boric acid solution corresponding to the most restrictive single failure in the safety injection system. The injection curve is shown on Figure 15.2-42. This corresponds to the flew delivered by one charging pump delivering its full contents to the cold leg header. No credit has been taken for the low concentration boric .acidthatmustbesweptfromtgsafetyinjectionlinesdownstreamofthe [$"Util MidC'[M"syh[h opm) to the reactor coolant loops. Odslia h N ~ e etmoea6t.ratsort boric acid ( X c, e.l. r k > s c* a ts.le 4s % A z T. y (a) The case studied is an initial total steam flow of 2?E lb/see at 1015 psia from one steam generator with offsite power available. This is the maximum l capacity of any single steam dump or safety valve. Initial hot standby conditions at time Zero are assumed since this represents the most pessimistic . initial condition. Should the reactor be just critical or operating at power at the time of a steam release, the reactor will be tripped by the normal overpower protection when power level reaches trip point. Ft110 wing a trip at power the RCS centains more stored energy than at no-load, the average coolant temperature is higher than at no-load, and there is appreciable energy stored in the fuel. Thus, the additional energy stored is removed via the cooldown caused by the steam line break before the no-load conditions of RCS temperature and shutdown margin assumed in the analyses are reached. The additional insertions proceed then in the same manner as in the analysis which assumes no. load condition at time fero. However, since the initial steam generator water inventory is 15.2-36 September 1987 Revision 3

DCPP UNITS 1 & 2 FSAR UPDAlt greatest at no-load, the magnitude and duration of the RCS cooldown are less for steam line breaks occurring at power. (5) In computing the steam flow, the Moody Curve f or f L/D. O is used. '(6) Perfect moisture separation in the steam generator is assumed. J 15.2.13.3 Results The results prnented are a conservative indication of the events that would occur assuming a secondary system steam release since it is postulated that all of the conditions described above occur simultaneously. FigureN15.2 43 M 'S ? 'lshoy the transients arising as the result of a steam release h 5 / having an initial steam flow of 228 lb/sec at 1015 psia with steam release from one safety valve. The assumed steam release is the maximum capacity of any single steam dump or safety valve. _In this case, safety injection is initiated automatically by low pressurizer pressure. Operation of one centrifugal charging pump is considered. Boron solution at =*aNa ,'d00$pmenterstheRCSprovidingsufficientnegativereactivityto += Y

1. W **4+4r=1 W S The reactivity transient f or the cases shown'on FigureK15.2-43 e@

" 2 " eehore severe than that of a f ailed steam generator safety or relief valve that is terminated by steam line differential pressure, or a failed condenser dump valve that is terminated by low pressurizer pressure. 'M le aiThe transient is quite conservative with respect to t e A cooldown since no credit is taken for the energy stored in the system metal other than that of the fuel elements or the energy stored in the other steam generators. Sir,ce the transient occurs over a period of about 5 minutes, the neglected stored energy is likely to have a . significant ef f ect in slowing the cooldown. EhKY l ~ 15.2.13.4 Conclusions The analysis has shown that the criteria stated earlier in this section are satisfied. the -^n'^ eM ^M ^*" a +a t****ra

• " ^^ * ! 4 " ' * ; c ' :

"?" tM t 15 Jess-than4.30# h .n_.~ -,._,.,... fu a 15.2.14 Sourious Doeratien of the Safety iniection System at Power 15.2.14.1 Identification of Causes and Accident Description Spurious S15 operation at power could be caused by operator error or a false electrical actuating signal. A spurious signal in any of the following channels could cause this accident. i 15.2-37 September 1987 Revision 3 A-38

i; w 1, t INSEFT A. 7 the limit DNER values will be met INSEET B l' exceed the-limit DNEF values. INSERT C For an accidental depressurinetion of the main steam system, the DNB~ design limite are'not exceeded. This case is -lees limiting than the rupture of a main'stean pipe case presented in .Section 16.4. 4 0 A-39

DCPP UNITS 1 & 2 FSAR UPDATE as they are coasting down following reactor trip. Therefore, upward flow through the core is neintained. The resultant heat transfer cools the fuel rods and cladding to very near the coolant temperature as long as the core remains covered by a two-phase mixture. This effect 6, is evident in the accompanying figures. The depressuritation transient for the limiting 4-inch break is shown on Figure 15.3-2. The extent to which the core is uncovered f or the same break is presented on Figure 15.3-3. The mautmum hot spot cladding temperature reached during the transient is 1288'F. including the effects of fuel densification as described in Reference 3. The peak cladding temperature transient for the limiting break size is shown on Figure 15.3 4. The core steam flowrate for the a-inch break is shown on Figure 15.3-5. When the mixture level drops below the top of the core, the steam flow computed in NOTRUMP provides cooling to the upper portion of the core. The rod film coef ficients f or this phase of the transient are given on Figure 15.3-6. Also, the hot spot fluid temperature f or the worst break is shown on Figure 15.3-7. Since a separate analysis was performed for DCPP Unit 1 a set of figures similar to those presented for the Unit 2 limiting break site can be found on Figures 15.3-14a through 15.3-14f. The core power (dimensionless) transient following the accident (relative to reactor scram time) is shown on Figure 15.3-8. The reactor shutdown time (4.4 sec) is equal to the reactor trip signal processing time (2.0 sec) plus 2.4 seconds for complete rod insertion. During this rod insertion period, the reactor is conservatively assumed to operate at rated power. Several figures are also presented for the additional break sites analyzed. Figures 15.3-9 and 15.3-10 present the RCS pressure transient for the 3-inch and 6-inch breaks, C respectively, and Figures 15.3-11 and 15.3-12 present the core mixture height plots for both breaks. The peak cladding temperature transient for the 6-inch break is shown on Figure 15.3-13. ho peak cladding temperature plot is shown for the 3-inch break since no core uncovery occurs (see Figure 15.3-11) and no clad heatup is computed. The snell break analysis was performed with the Westinghouse ECCS Snell Break Evaluation I Mode 1 approved for this use by the Nuclear Regulatory Commission in May 1985. 15.3.1.4 Conclusions Aqalyses presented in this section show that the high-head portion of the CCCS, together with the accumulators, provides sufficient core flooding to keep the calculated peak cladding temperatures below required limits of 10 CFR 50.46. Hence adequate protection is afforded by the ECCS in the event of a small break LOCA. l-15.3.2 Minor Secondarv system Pioe Breaks 15.3.2.1 Identification of Causes and Accident Description included in this grouping are ruptures of secondary system lines which would result in steam release rates equivalent to a 6-inch-diameter break or smaller. 15.3-4

• * # ' '** * ' ' " 7 ' ' " " ' " *

-40

DCPP UNITS 1 & 2 FSAR UPDATE 15.3.P.2 Analysis of Effects and Consequences Minor secondary system pipe breaks must be accommodated with the f ailure of only a small fraction of the fuel elements in the reactor. Since the results of analysis presented in Sectirsn 15.4.2 for a major secondary system pipe rupture also meet these criteria, separate analyses for minor secondary system pipe breaks is not required. The ar.alyses of the more probable accidental opening of a secondary system steam dump, relief, or safety valve is presented in Section 15.2.13. These analyses are illustrative of a pipe break equivalent in size to a single valve opening. 75.3.7.3 Conclusions The analysis presented in Section 15.4.2 demonstrates that the consequences of a minor secondary system pipe break are acceptable since a departure from nucleate boiling ratto (DNBR) of less than'4rMedoes not occur even for a more critical major secondary system pipe 7 L rL do,y [n,> wlstea 15.3.3 Inadvertent loadina of a Fuel Assembly Into an Imorecer Position 15.3.3.1 Identification of Causes and Accident Description fuel and core loading errors such as can arise from inadvertently loading one or more fuel assemblies into improper positions, loading a fuel rod during manufacture with one or more pellets of the wrong enrichment, or loading a full fuel assedly during manufacture with pellets of the wrong enrichment will lead to increased heat fluxes if the error results in placing fuel in core positions calling for fuel of lesser enrichment. The inadvertent loading of one or more fuel assemblies requiring burnable poison rods into a new core without burnable poison rods is also included among possible core loading errors. Any error in enrichment, beyond the normal manufacturing tolerances, can cause power shapes that are more peaked than those calculated with the correct enrichments. The incere system of movable neutron flux detectors that is used to verify power shapes at the start of 11fe is j capable of revealing any assembly enrichment error or loading error that causes power shapes ) to be peaked in excess of the design value. To reduce the probability of core loading errors, each fuel assembly is marked with an identification number and loaded in accordance with a core loading diagram. During core loading, the identification number will be checked before each assembly is moved into the i core. Serial numbers read during fuel movement are subsequently recorded on the loading I diagram as a further check on proper placing af ter the loading is completed. O 1 l l 15.3-5 A-41 i j

DCPP UNITS 1 & 2 ISAR UPDAT[ l (4) The core temperature is reduced.and decay heat is removed for an extended period of time, as required by the long-lived radioactivity' remaining in the core. 15.4.2 Ma_ior Secondarv system Pipe Ruoture

• ~

Two pujor secondary system pipe ruptures are analyzed in this section: rupture of a main - steam line and rupture of a main feedwater pipe. The time sequence of events for each of these events is provided in Table 15.4-B. 15.4.7.1 Ruoture of a Main Steam Line 15.a.?.1.i ' identification of Causes and Accident Description The steam release arising from a rupture of a main steam pipe would reruit in an initial . increase in steam flow that decreases during the accident as the steam pressu'e falls. The energy removal f rom the RC5 causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coef ficient, the cooldown results in a reduction of core shutdown margin. If the most reactive RCCA is assumed stuck in its fully withdrawn - position af ter reactor trip,' there is an increased possibility that the core will betone - - critical and return to power. A return to power following a steam pipe rupture is a potential problem mainly because of the high power peaking factors that exist assuming the' most reactive RCCA to be stuck in its fully withdrawn position. The core is ultimately shut down by the boric acid injection delivered by the 515. The analysis of a main steam pipe rupture is performed to demonstrate that the following criteria are satisfied: (1) Assuming a stuck RCCA, with or without offsite power, and assuming a single f ailure in the en2ineered safety features (ESF) there is no consequential damage to the primary system and the core remains in place and intact. (2) Energy release to containment f rom the worst steam pipe break does not cause failure of the containment structure. . Although DNB and possible cladding perforation following a steam pipe rupture are not necessarily unacceptable, the following analysis, in fact, shows that no DNB occurs for any rupture assuming the most reactive assembly stuck in its fully withdrawn position. The following functions provide the necessary protection against a steam pipe rupture: A 15.4-6 A-42

l DCPP UhlTS 1 & 2 FSAR UPDATE (1) 515 actuation from any of the following: .( a) Two-out-of-four low pressurizer pressure signals (b) High differential pressure signals between steam lines (t) High steam itse flow in two main steam lines (one-out-of-two per line) in coincidence with either low-low RCS average temperature or low steam line pressure in any two lines (d) Two-out-of-three high containment pressure. (2) The overpowr reactor trips (neutron flux and AT) and the reactor trip occurring in conjunction with receipt of the safety injection signal. (3) Redundant isolation of the main feedwater lines: sustained high feedwater flow would cause additional coeldown. Therefore, a safety injection signal will rapidly close all feedwater control valves, trip the main feedwater pumps, and close the feedwater isolation valves that backup the control valves. (4) Trip of the fast acting main stese line isolation valves on: (See Fig. T.2-1 I UI and Technical Specification Table 3.3-5) (a) High steam flow in two main steam lines in coincidence with either low-low T or low steam line pressure in any two lires g (b) High-high containment pressure. The fast-acting isolation valves are provided in each main steam line and will fully close within 10 seconds of a large steam line break. For breaks downstream of the isolation valves, closure of all valves would completely terminate the blowdown. For any break, in any location, no more than one steam generator would blow down even if one of the isolation valves fails to close. A description of steam line isolation is included in Chapter 10. Steam flow is measured by monitoring dynamic head in nortles inside the steam pipes. The norries that are of considerably smaller ciameter than the main steam pipe are located inside the containment near the steani generators and also serve to limit the maximum steam flow for any break further downstream. 15.a.P.1.2 Anaivsis of Effects and Conseovences The analysis of the steam pipe rupture has been performed to determine: k

15. 4 -T A-43 o

r-DCPP UN!15 1 & 2 FSAR UPDAl[ (1) The core heat flux and RCS temperature and pressure resulting f rom the cooldown f ollowing the steam line break. The6$m %{+- code has been used. y L LCTIQW W (~ (2) The thermal and hydraulic behavior of the core following a steam line break. . A detailed thermal and hydraulic digital-computer code, THINC , has been used to determined if DNB occurs for the core conditions computed in (1) above. The f ollowing conditions were assumed to exist at the time of a main steam line break accident. (1) End of life (EOL) shutdown margin at no-load, equilibrium menon conditions, and the most reattive assembly stuck in its fully withdrawn position: Operation of the control rod banks during core burnup is restricted in such a way that addition of positive reactivity in a steam line break accident will not lead to a more adverse condition than the case analyzed. (2) The negative moderator coefficient corresponding to the EOL rodded core with the most reactive rod in the fully withdrawn position: The variation of the coefficient with temperature and pressure has been included. The k versus temperature at hcorrespondingtothenegativemodera [ r r ap l temperature coefficient % plus the Doppler temperature effectgis shown on A Figure 15.2 41. The effect of power generation in the core on overall f reactivity is shown on Figuref i5. 6.nd 15.4-64 The core properties associated with the sector nearest the af fected steam generator and those associated with the remaining sector were conservatively combined to obtain average core properties for reactivity feedback calculations. Further, it was conservatively assumed that the este power distribution was uniform. These two conditions cause underprediction of the reactivity feedback in the high-power region near the stuck rod. To verify the conservatism of this method, the reactivity as well as the power distribution was checked for the statepoints shown in Table 15.4-9. These core analyses considered the Doppler reactivity from the high fuel temperature near the stuck RCCA, moderator feedback from the high water enthalpy near the stuck RCCA, power redistribution and non-uniform core inlet temperature effects. For cases in which steam generation occurs in the high flux regions of the core, the ef fect of void it.rmation was also included. it was determined that the reactivity employed in the kinetics analysis was always larger than the true reactivity for all statepoints in Table 15.4-9, verifying conservatism; i.e., underprediction of negative reactivity feedback from power generation. 15.4-8 S'pt'aber 1987 8' vision 3 A-44

DCPP UNITS 1 5 2 FSAR UPDAlt Di q. A (2) Minimum capability for injection of high concentration boric acid (Hr;996 ppm) solution corresponding to the most restrictive single failure in the SIS. The characteristics of the injection unit used are shown on Figure 15.2 42. This corresponds to the flow delivered by one charging pump delivering its full flow to the cold leg header. No credit has been taken for the low concentration of beric acid tnat must be swept from the safety injection lines y Le ta#.e. ".;;i '"") isolation va,l}ves prior to the we.re, s r :. - e o n..,8 t w.. - delivery of highly concentrated boric acid to the reactor coolant loops. Mh-+' K downstream of the ;e:en n + Meet 4e; M;; :11;cd for-4n-tM en;MM The modeling of the 515 in ( is described in Reference N. [ For the cases where offsite power is assumed, the sequence of events in the 515 is the following: After the generation of the safety injection signal (appropriate delays for instrumentation, logic, and signal transport included),theapprgoriatevalvesbegintooperateandthehigh-headinjection pump starts. Inhseconds,thevalvesareassumedtobeintheirfinal position and the pump is assumed to be at full speed. The volume containing

t. y n...,r.,6c..

r..- t ~.. t s t sh < 1 e s' the M,,:. is swept before thegGAG2-ppm boron reaches the core. This 1/t L delayisinherentlyincludedipthemodeling. In cases where offsite power is nei available, an additional b second delay is assumed to be required to start the diesels and to load the necessary safety injection equipment onto them. That is, af ter a total of M seconds f ollowing an $15 signal, the $15 \\; N is assumed to be capable of delivering flow to the RCS. (4) Four combinations of break stres and initial plant conditions have been considered in determining the core power and RCS transients: (a) Complete severance of a pipe outside the containment, downstream of the steam flow measuring norrie, with the plant initially at no-load conditions, full reactor coolant flow with offsite power available (b). Complete severance of a pipe inside the containment at the outlet of the { steam generator with the plant initially at no-load conditions with ] offsite power available ) (c) Case (a) above with loss of of f site power simultaneous with the f 1 initiation of the safety injection signal. Loss of offsite power results j in coolant pump coastdown. J (d) Case (b) above with the loss of offsite power simultaneous with the initiation of the safety injection signal. t 15.4-9 A-45

l' DCPP UNITS 1 & 2 FSAR UPDATL (5) Power peaking factors corresponding to one stuck RCCA and non-uniform core inlet coolant temperatures are determined at EOL. The coldest core inlet temperatures are assumed to occur in the sector with the stuck rod. The power a peaking factors account for the effect of the local void in the region of the stuck control assembly during the return to power phase following the steam line break. This void in conjunction with the large negative moderator coefficient partially offsets the effect of the stuck assembly. The power peaking factors depend on the core power, operating history, temperature, pressure, and flow, and thus er dif. erent for each case studied. ny'st D y f(OW N) +i ? ac. m p ye The values used f og Wie ;' a iver steamline break accidenth-d,;;; are i given in Table 15.4-9. ": t":: :::::

::":^:d :. t'; h;'; :' t:t-24 -:::tw-a+0 ant-tmsurer-he-f ourth case is Y

1 dennel-feeter+r4e*: ;;;:- 4est4ewar* r elat.ivt in the denattur.e_hnntnucleate bniling talin.DNER). The _ N

n: ;;r:=':r; ;;-d ':: :::' :' tM t"ee ;ese; ;s,.. a yd te M.i; i

r . determined ina-theasser ma teastarstarsalysis,---Tive-time point +-ere-used Y

n. 9 em All the cases above assume initial hot standby conditions at time zero since this represents the most pessimistic initial condition. Should the reactor be just critical or operating at power at the time of a steam line break, the reactor will be tripped by the normal overpower protection system when power level reaches a trip point. Following a trip at power the RCS contains more stored energy than at no-load, the average coolant temperature is higher than at no-load, and there is appreciable energy stored in the fuel. Thus, the additional stored energy is removed via the cooldown caused by the steam line break before the no-load conditions of RCS temperature and shutdown margin assumed in the analyses are reached. After the additional stored energy has

~ been removed, the cooldown and reactivity insertions proceed in the same manner as in the analysis which assumes no-load condition et time zero. However, since the initial steam generator water inventory is greatest at no-load, the magnitude and duration of the RCS cooldown are less for steam line breaks occurring at power. (6) In computing the steam flow during a steam line break, the Moody Curve for fL/D = 0 is used. The Moody Multiplier is 1 with a discharge at dry saturated steam conditions. (7) Perfect moisture separation in the steam generator is assumed. The assumption leads to cor'servative results since, in fact, considerable water would be discharged. Water carryover would reduce the magnitude of the temperature decrease in the core and the pressure increase in the containment. 15.4-10 September 1981 Revision 3

4 L ?' 1. ItiSEhi b A tots) of fivat tinm points are g precented. Thie case is celected on i the barie of hot channel factore, core b power. and reactor coc] ant pressure. The other'three esser are lees severe re l.s t i ve to departure from' nucleate be,iling ratio- (DITBli)

The core param-tere.used for each t>f the four canes correspond to values. determined fret the rdepective transient analysic O

l O A-47

DCPP UNITS 1 & 2 ISAR UPDF.1E 15.4.2.1.3 Results i The results presented are a conservative indication of the events that would occur assuming a steam line rupture since it is postulated that all of the conditions described above occur simultaneously. X Figure [15.4-65 ;d l',.7-46 show the RCS transient and core heat flux following a sein steam o pipe rupture (complete severance of a pipe) outside the containment, downstream of the flow measuring nozzle at initial no load condition (Case A). The break assumed is the largest break that can occur anywhere catside the containment either upstream or downstream of the isolation valves. Of f site power is assumed available such that full reactor coolant flow exists. The transient shown assumes an uncontrolled steam release from only one steam generator. Should the core be critical at near tero pow *r when the rupture occurs, the initiation of safety injection by high differential pressure between any steam line and the remaining steam lines or by high steam flow signals in coincidence with either low-low RCS temperature or low steam line pressure will trip the reactor. Steam release from more than one steam generator will be prevented by automatic trip of the fast action isolation valves in the steam lines by the high steam flow signals in coincidence with either low-low RCS temperature or low steam line pressure. Even with the failure of one valve, release is ilmited to no more than 10 seconds for the other steam generators while the one steam generator blows down. The steam line isolation valves are designed to be fully closed in less than 5 seconds af ter receipt of closure signal with no flow through them. With the high flow existing during a steam line rupture the valves will close considerably faster. (h ThesteamflowonFigures15.4-65through15.4-)s,representssteamflowfromthefaulted [ steam generator only. In addition, all steam generators were assumed to discharge through the break for the first 10 seconds. bl h o As shown on Figuret15.4-N ed l',.:;O, the core attains criticality with the RCCAs inserted (with the design shutdown assuming one stuck assembly) before boron solution at p'h000 ppm uw ar enters the RCS from the 515. The delay time consists of the time to receive and actuate the safety injection signal and the time to completely open valve trains in the safety injection At this stage a further The safety injection pg'ngs are then ready to deliver flow. lines. l delay time is incurred before Mi49G-ppe boron solution can be injected to the RCS due to the h low concentration solution being swept from the safety injection lines. A peak core power well below the nominal full power value is attained. The calculation assumes the boric acid is mixed with and diluted by the water flowing in the RCS prior to entering the reactor core. The concentration after mixing depends on the l relative flow rates in the RCS and in the 515. The variation of mass flowrate in the RCS due to water density changes is included to the calculation as is the variation of flowrate 4 15.4-11 A-48

DCPP UNITS 1 & 2 FSAR UPDAff from the SIS and the accumulator due to changes in the RCS pressure. The SIS flow calculation includes the line losses in the system as well as the pump head curve. The accumulators provide an additional source of borated water after the RCS pressure decreases to below h sia. The integrated flowrate of borated water from both the accumulators and 7 the $15 for each of the four rases analyzed is shown on Figureg15.4 73 W E' . g g g figuref 15.4 b.ni G. -on show, Case g. a steam line rupture at the exit of a steam A generator at no-load. The sequence of events is similar to that described above for the rupture outside the containment except that criticality is attained earlier due to more rapid cooldown and a higher peak core average power is attained, p 'h Figures 15.4-Wthrough 15.4-Wshow the responses of the salient parameters for Cases C and g which correspond to the cases discussed above with additional loss of offsite power at the njec_ tion si The $15 delay time b k seconds to N E N [::gnal is generated.;:t th:-safety-injection -pumps-to-firil-epeeth' In I time the safety M-t: start the diese 4 case criticality is achieved later and the core power increase is slower than in the similar case with offsite power available. The ability of the emptying steam generator to extract ~ heat from the RCS is reduced by the decreased flow in the RCS. For both these cases the peak core power remains well below the nominal full power value. It should be noted that following a steam line break only one steam generator blows down completely. Thus, the remaining steam generators are still available for dissipation of decay heat after the initial transient is over. In the case of loss of offsite power this heat is removed to the atmosphere via the steam line safety valves that have been stred to cover this condition. l 15.4.2.1.4 Conclusion A DNB analysis was performed for the three cases most critical to DNB. Five points from each case were examined. It was found that all cases had a minimum DNBR greater than HKN-il e. /..,,./ y W./a e 15.4.2.7 Maior Ruoture of a Main Feedwater pipe 15.4.7.7.1 Identification of Causes and Accident Description A major feedwater line rupture is defined as a break in a feedwater pipe large enough to prevent the addition of sufficient feedwater to the steam generators to maintain shell-side fluid inventory in the steam generators. If the break is postulated in a feedline between the check valve and the steam generator fluid from the steam generator may also be discharged through the break. Further, a break in this location could preclude the subsequent addition of auxiliary feedwater ( AFW) to the af fected steam generator. (A break upstream of the feedline check valve would af fect the nuclear steam supply system [NS$$) only l as a loss of feedwater. This case is covered by the evaluation in Section 15.2.8). 15.4-12 A.49 september 1986 Revision 2 i

,k. g i 'I11SEFT E' and load the necescary eafety injection equipment onto them, for the appropriate valver to reach their final position, and for the high-head-injection pump t o reach full cpeed. e e l' 9 A-50

.-)(' 3, i

s, s

f CPP.UN1151 & 2 FSAR UPDATE r TABLE 15.2-1 Sheet 12 of 13 .g.. Accident Event Tine. see Accidental Depressuri-ration of the Reactor Coolant System Inadvertent opening of one RCS safety valve 0.0 Reactor trip 28.3 Minimum DNBR occurs 30.5 Accidental Depressuri-zation of the Ulitf1 V'ti t f Main Steam System Irsadvertent opening of one main steam safety i S or relief valve 0.0 6 I' .0 Pressurizer empties 10 g27 1 2 Je t' 20;900 ppm boron reaches RCS loops

193,

.I A S. 1 4 A-51 l

7193-47 l. l.07 ZERO POWER,1000 P51 A EnD OF Lif f RODDID .s CORI WITH Oh[ RCC A STUCK FULL 0U1 i.o6 ,/ \\ i.05 \\ / \\ / uun i e g i.ou \\ / \\ / = \\ / t; I.03 \\,/ s 'N / \\ c i.02 g \\ own 2/ g a s / / k I \\.-s 1.01 7 3 E /\\ s i.oo / N / 'N / 0.9s s / O.96 s / 's / (37 i I l 'i i \\ 200 250 300 SFO 400 450 \\ 500 55 0 CORE MERAGE TEMPERATURE (OF)\\'s / \\ / \\ / s Figure 15.241 Variation of K,g with Core Temperature ,-( ht:i t 6 ' \\ br 2 -4 1 ng.v Iy A-52

I ZB0 30WB, ' 50 'S A ;30;; ;1011D C0;E ~ WITIl 0.NE RCCA STiiCE rl!LL OlIT 1.01 N\\ 1.03-N 'N\\ 0 b 1.02-0 \\ \\ L \\\\ [. \\\\ 2 1.01-E \\, 0 3 O L 1-F J D 2 0.99 - 0.98 i i i i i i 200 300 400 500 600 CORE AVERAGE TEMPERATURE (DEG-f) FIGURE 15.2-41 Variation of Keff with Core Temperature A-53

p 7193 141 2400 2200 o- / / 1800 / / / ,/ 1600 / / n 1400 w ./ \\ ~ 1200 / D -\\/ g 1000 ,/'\\ /' \\ / g / \\ 800 a '\\ 600 j ,i \\ 400 z_ N / \\ / \\ 2% \\ \\ \\ I I 0 O 100 200 300 400 500 6'00 700 SAFETY INJECTION FLOW (GPM) \\ \\ FROM ONE CENTRIFUGAL CHARGING PUMP \\ \\ m Figure 15.2-42 Safety injection Curve, Main Steam Depressurization ,.Ga v c. O n ' A-54 2"42 r, 's ) 5

E 'o ' g. M-g. IR - 1E-7 i 1G-w Eg 125 - w E y 1E-o. g. p. g. O i e i i i i E 5 W SAFETYINJECTIONFLOW(GPM) l FROM ONE CENTRIFUGAL CHARGING PIMP 1 l l~ 1 l. FIGUEE 10.2-42 Safety Injection Curve, 1-Rain Steatu Ilepre s e uri stion f-A-55

7893-32 3000 \\ W l l l l \\ j @ 20.000 PPM BORON RE ACHES g ga 2000 LOOPS Al 192 SEC o u"5 e = m S 5 ' 1000 WM wg 0 / 700 5d 45}N 600 =w wwo 5$ 009 \\ 400 / \\ / / \\ / / 2.5 ? C-y ',' -x 0 $ 21 -2.5 ,/ x I I -5.0 0 / 100 200 '300 400 500 7 TIME (SECONDS) / / i ,/ l / / e Figure IS.2-43 Transient Response for o Steam Line Break Equivalent to 228 lbs./Sec of 1015 PSIA with Outside Power Available (Unit I) 1 43 nt i-g' y ,a "t } "~$/ I A-56

p F 2 ECD.< l*- n l-- Q 2000. ld- ' 1503. - E!iebe. . Ww 500. B. .600. W. 503. g 20 403. w" b - 503, 223. 2-2,300 PPM BORON REACHES I _"~ LOOPS AT 328 SEC _N g*. PE o ~ $.W

• t'.-

-1. -2 B. 123. 222. 522. dca. 520. 600. TIME ISECl FIGURE 15.2-43 DIABLO CANYON TRANSIENT RESPONSE FOR A STEAM LINE BREAK EQUIVALENT TO 228 LBS /SEC AT 1015 PSIA WITH OUTSIDE POWER AVAILABLE A-57

7193-33 3000 l I I l l ,_. m 5$ 20,000 PPM BORON REACHES y @ _ 2000 LOOPS AT 201 SEC v=5 a m N 1000 =~ 0 700 / / 600 5e \\ / /

a ; 500

/ xg u t *. 85 400 g / \\ / 300 2.5 f /y -b2 0 r; 1 co / a li = -2.5 / - / I I ( l I -5.0 i 0 100 200 300 400 500 600 1 \\ TIE (SECO DS) p / 1 / l. / l Figurejl.2-44 Transient Response for o Steam Line Greek Equivdent to 228 Lbs/Sec et 1015 PSIA with Outside Power Available (Unit 2) N \\ -__-___-___-_-______u

l. DCPP UNITS 1 & 2 FSAR UPDATE j TABLE 15.4-8 Sheet 1 of 3 TIME SEQUENCE OF EVENTS FOR MAJOR SECONDARY SYSTEM PIPE RUPTURES Accident Event Time. sec i nitI2 IJnitil Major Steam Line l i 1. Case A Steam line ruptures DfC 0 Criticality attained 19 ' 5' 6-Pressurizer empty 15 ,+ 1 5l 2 re ; e 20iO90 ppm boron reaches loops 33 l 7"I 23 i 2. Case B Steam line ruptures 0 o (O Criticality attained 14 35 13l Pressurizer empty 19 /t 18 2 3c c i i. 20r000 ppm boron reaches loops 35 7e 25 3. Case C Steam line ruptures 0 D 0 If Criticality attained 2I ( #l Pressurizer empty 15 /5 16 2 pc \\ 20,000 ppm boron reaches loops 43, '7+ 3p !i l 'i 0 4. Case D Steam line ruptures JO O O +8 15 Criticality attained l7 P{g,s,surizerempty 39 i6 1, 9, 207000 ppm boron reaches loops WT 73 35 \\ l\\ \\ !l i A-59 September 1986 Revision 2

6 3 7 0 6 2 3 t 9 7 9 0 9 Q i 3 4 5 1 f n o 5U 3 8 4 5 1 1 0 t 2 8 8 0 6 7 e t 9 6 7 0 9 e i 3 4 5 1 h n S U 9 9 5 4 3 5 2 6 5 2 0 7 7 t 0 8 0 0 6 i 4 4 7 1 n U 3 41 2 2 1 8 t 5 3 3 0 6 0 t i 0 8 9 0 7 \\ n n 4 4 6 1 i U \\ o s S P 4 \\ I 2 9 2 5.\\' 1 S e Y m t 1 3 9 0 8 5 L i i 2 9 9 0 4' A T n 4 4 7 1 N U A 9 A 1 6 7 5 B 31 E N e 0 2 6 0 7 T D s t 2 9 9 0 4 A a i 2 4 7 1 D K C n P A U N C E x 8 R 1 3 1 2 5 R B 2 A 9 5 4 'N 9 0 8 4 S M t 3 9 4 0 3 A i 4 4 8 1 F 4 E n 5 T U 2 1 S 2 / 3 o E N 1 3 6 6 5 L I 1 B t 9 5 8 0 6 2 A D i 3 9 6 0 3 S T E n 4 4 8 1 T S .U / O I U N 6 / U S 2 2 8 5 7 R P E t 3 4 9 0 6 0 P I i 4 9 6 0 3 C L n 4 4 8 1 D M U A R 1 ' 8 A 1 8 4 3 1 P t 4 7 7 0 6 5 E i 5 9 0 0 2 R n 4 4 9 1 D ,U Cy .F - oF r' rt" e e. pd, pr mer mo et o et t ct t c ea e t nr t s ene e l on l g a G nce nn i i g i i s r n p Y om l l l r et a ea ~ e sce s n e t set se r e ess er u m v v s x t a od o s w u e r rt e rt e o l s a o t o r l f P t ec t e p I cre cr t e auf au S S a m 0 etf et C C e i Raa Ra R R H T >&o

4 7 4 0 2 o 8 8 4 0 2 0 3 t 5 8 4 0 1 7 3 4 6 1 i f n o 5U 5 4 2 6 8 9 7 5 1 t 8 7 7 0 7 2 e t 3 2 8 0 2 e i 3 4 4 1 1 h n S U 8 7 3 5 5 5 2 7 6 5 0 2 9 t 6 0 9 0 1 4 i 3 5 7 1 n U 6 1 8 4 41 t 0 1 3 0 8 0 t i 4 5 2 0 0 n n 3 4 5 1 1 i U o .l~ P 5 2 5 5 6 5 em t 2 2 9 0 1 1 i i 7 1 4 0 1 4 \\.s T n 3 5 8 1 U \\ 5 8 7 3 9 6 \\ 31 E e 8 0 3 0 9 0 T s t 4 7 7 0 8 A a i 3 4 5 1 D C n P U g U 5 \\ 9 4 9 4 R 2 A 9 5 6 9 0 2 7 S t 7 1 7 0 1 3 \\ i 3 5 8 1 F 4 5 U \\-g n 2 1 2 3 7 E 1 8 8 9 L \\ 1 B t 7 8 8 0 8 5 A i 7 1 9 0 3 S T n 3 5 8 1 T _U I N 7 U 2 7 5 8 5 5 P t 6 3 8 0 9 7 P i 8 2 9 0 3 C n 3 5 0 1 D U 1 1 2 4 8 9 2 1 t 1 6 3 0 8 5 i 9 2 0 0 2 n 3 5 3 1 U 1 F - oF rt" r e e, s\\ pd pr mer mo et o et t ct t c ea e t nr t s ene e l on l g a nce nn i i g ii s r n p om l li r et a ea e sce sm e t set se r g f e ess er u m v v s t t a od o s w u e r rt e rt e o l s a o t o r l f P t ec t e p f cre cr t e auf au S S a m etf et C C e i R aa Ra R R H T , 3 3

f n o 5U 9 3 5 5 5 5 5 1 t 1 4 7 7 3 7 e t 8 2 7 1 8 e i 2 5 8 h n S U 2 5 3 9 1 2 3 7 8 0 5 5 t 0 2 9 2 6 i 3 5 9 n U 2 41 1 9 8 5 5 t 2 8 4 2 4 7 t i 1 2 0 2 5 / n n 3 5 0 s U 1 \\ io P 7 2 1 9 1 9 8 5 \\ e f m t 2 8 2 4 7 0 /i i 2 2 1 2 5 T n 3 5 1 U 1 0 5 1 4 7

2. \\ 'S.

31 E e '5 9 0 0 6 7 T s t 4 2 7 3 3 A a i 3 / 5 2 D C n 1 P U, U 7 4 9 2 / 8 3 7 R 2 / A 9 0 9 4 2 6 5 S t 5 2 6, 3 3 F 4 i 3 5 2 / n 1 5 U 2 1 2 /, E 1 L 6 9 9 5 1 1 I. B t 6 8 9 3 6 2 A i 5 2 5 3 3 S T .n 3 5 3 T .U 1 I N U 2 5 9 3 6 8 P t 5 9 9 0 5 5 P i 7 2 6 4 2 C n 3 5 4 0 U 1 1 7 s 1 5 2 5 2 t 0 0 2 7 5 0 N i 9 3 1 4 2 n 3 5 7 _U 1 \\ s \\ \\ F \\ - oF r' \\ rt* e e. pd, pr mer mo et o et t ct t c ea e t nr t s ene e l on l g a nce nn i i g ii s r n p om l li r et a ea e sce sm e t set s e r e ess er u m v v s x c a od o s w u e r rt e rt e o l s a o t o r l f P t ec t e p f cre cr t e auf au S S a m etf et C C e i Raa Ra R R H T p O" )3

ll TABLE 15.4-9 l CORE PARAMETERS USED IN STEAM BREAK DNS ANALYSIS l l Case B, Time Point Parameter 1 2 3 4 5 Reactor vessel inlet temper-324.6 324.6 324.6 324.7 325.8 ature to sector connected to affected steam generator, 'F Reactor vessel inlet temper-422.7 422.6 422.6 422.5 422.5 ature to remaining sector, 'F RCS pressure, psia 602.9 603.5 604.3 605.2 606.6 RCS flow, % 100 100 100 100 100 Heat flux, % 19.0 19.0 19.0 19.0 19.0 Time, sec 266.2 268.2 270.2 272.2 274.2 ,.,,- c i T s s r 0738v.1D/D81088 A-63

7i93-56 i l O (, / V'{,lOV \\ '\\ -2000 g \\ j \\ / E\\ / r \\ M '\\ \\ ./ S \\ f ,/ g \\ o \\ / / -1000 -N e y e s g / 'N j \\ E / \\ / S a E / \\ <' v / x r I / I'N l l 0 10,,/ 20 30 40 50 60 0 ,/ POWER (FRACil0H 0F 3350 MWt) s / / / / / \\ o. \\ 's / Figure 15.4-63 Variation of Reactivity with Power of Constant Core Averoge Temperature (Unit 1) a j / / / A-64

7193-57 \\ \\ a- -2000 / / _g } / y l / 1 y / / y S i u o -1000 5 \\ g \\,' x'\\ m 5 / / N / \\ / l ', N! / ~ O y 0 0.10 0.20 0.30, 0.40 0.50 0.60 POWER (FRACTION OF S423 MWt) \\ \\ \\ / \\ l- / \\ / \\ / \\ / \\ / \\ / .\\ / \\ \\ \\ o Figure 15.4-64 Variation of Reactivity with Power of Constant Core Average Temperature {UrM ?) e. o V ' 4 )., n t* n / . ; lca t i g 'I ~ l,f A-65

9 Ac -2000 Y E $w M i.u' o -1000 5 E o. c5 d 5 / U f. 5 / / / l 0 0 0.10 0.20 0.30 0.40 0.50 0.60 POWER (FRACTION OF 3423 MWt) io-Figure 15.4-64 Variation of Reactivity with Power of Constant Core Average Teniperature o-A-66 h

7103-67 1 600 i I I l-1 I I I g = { 500 t-m 5 g 400 "8* 1 300 j 3000 gg PRESSURIZER EMPTIES AT 15 SEC 3a 8 i 2000 "w 8E e = 1000 m E Eh. \\ o N 3.0 N INillAL STEAM FLOW IS 3C82 LB/SEC FROM THE \\, F AULlED STEAM GENERATOR ( AND 927E LB/SEC \\ FROM THE INTACT STEAM GENER ATOR) 2.5 l3 E x / j' l;E E 2.0 - EE w" / dw ug "J _ I5 - i my tw x: w o W H / gi W5 / FAULTED STEAM GENERATOR ONLY u - a g 1.0 l O.5 / .,t s -------- 9 25 3; 20,000 PPM BORON RE ACHES LOOPS A123 SEC. t4 > s' 0

5 t' s

~ l i I i l i I /g = -2.5 0 25 50 75 100 125 150 175 200 TIME (SECONDS) / ~ / Figure 15.4-65 Transient Response to Steam Line Break Downstream of F Nozzle With Safety injection and Offsite Power (case a), Unit i W !" d A-67 I,. G e < w ,g_ yl & I

620. ui. ECO. 5, gagdem.

500, 220.

2507. .- 7 5 2 2000* PRESSURIZER EMPTIES AT 14 SEC g 'h W O 1000.

• E 500.

B. .5' ,h .25 ' Nl .2 E .16 - g .1 - "$.es. 8. 3~ INITIAL STEAM FLOW IS 2900 (R/SEC FROM THE FAULTED STEAM GENERATOR AND 8700 LB/SEC (2900 PER $G) -g,,, FROM THE INTACT STEAM GENERATORS 85 9 U INTACT STEAM GENERATORS E8 wt "O \\ JAULTED STEAM LENERATOR ONLY Q .2 (. B. 2 ~ 2300 PPM SONON REACHES CORE AT 84 $tt y ,N E$ 0 ^ Es ry -1 2 2.

50. 182. ISB. 2ft. 250. 502. 550. 488. 458. 588.

TIME ISCC) e FIGURE 15.4-65 DIABLO CANYON TRANSIENT RESPONSE TO STEAM LINE BREAK DOWNSTREAM OF FLOW MEASURING DEVICE WITH SAFETY INJECTION AND OFFSITE POWER (CASE A A-68 w

l l 7193-63 s 600 I I I I I I I x \\ NC 500 l m@"b s ^E$g 1600 %5w ow 300 3000 0-PRESSURIZER EMPilES AT 15 SEC ,/ 5\\5 a g 2000 j t W E%s1000 20 N wa \\ / O.O. 3.0 INil: AL STEAM FLOW M92 LB/SEC FROM FAULTED STEAM GENERATOR ( AND 9276 LB/SEC FROM THE INTACT J1EAM GENERATORS) l lC O j I E lg = 2.0 - j/ 5E w" dw QK /.' 5 i.5 - gg w p

• ux*

/ W5 52 8 6 ti d 1.0 / FAULTED STEAM GENERATOR CHLY b i / 0$ - / / / / n r--~~~-- /22.5 l r.x 20.000 PPM BORON REACHES LDOPS AT 33 SEC.

- c p%

0 ^ i e0 1 "k-2.5 I I I I I I I / 0 25 50 75 100 125 150 175 200 / TIME (SECONDS) / v lys 'U'g,t, ?, t t Figure 15.4-66 Transient Response to Steam Line Break Downstream of Flow Measuring Nozzle With Safety injection and Offsite Power (case a), Unit 2 i A-69 i i

'i l ) sta. NO'502. 'I Eb 3I ' 402. g agg a-w Ece.. 220. 2500. PRES $URIZER [MPT![$ AT 16 $[C

pggg, 1500.

R,m. 5 " E"' ' ct. S 9. .5 .25 dl- .2 . is. gg "E .es. 8. 7 INITIAL STEM FLOW IS M00 (B/SEC FROM THE $h-2.5 FAULT [D STEM GENERATOR AND 2900 LB/SEC (M6 PER $G) g FROM THE INTACT $T[AM GENERATORS 2. is i.5 35, INTACT STEAM CENIRATDRS s. f .5, FAULTLD STLAM GINfRATOR ONLY s / e. 2- { 2300 PPM BORON REACHE5 CORE AT 83 $EC 5 0 E5 E -1 .I e. te. 122. 158. Ite. 252, 532. 550. ace. Tlat estCs ~ FIGURE 15.4-66 DIABLO CANYON TRANSIENT RESPONSE TO STEAM LINE BREAK AT EXIT OF STEAM GENERATOR WITH WATER INJECTION AND OFFSITE POWER (CASE B) A-70

l 1 7893-64 \\ d I I I I I I I s ,r 'N Rs _. 500 '\\ E E a. io y e d 5 400 g-s 300 E

• 3000 PRESSURIZER [MPTIES AT 18 $!C s\\G 8 *'x 2000 u=EN

,b 000 / N5 \\ 0\\ A 3.0 INIT I AL STE AM F LOW 15 11253 LB/SEC FROM , F AULTED STEAM GENERATOR ( AND 2964 LB/SEC 'f ROM INT ACT STEAM GENER ATOR) 2.5 N lG S X lE, g 2.0 x EE m" dw 28 "i2 1.5 s hw F AULTED STE AM GENERATOR ONLY me 5-EE "5 1.0 u-m y E s 0.5 \\ \\ s \\ j \\ 20.000 PPM BORON REACHES LOOPS AT\\25 SEC Ea / \\ Er 0 ^ U5 'N OE l l l l l l l c/ t -2.5 ~ / 0 25 50 75 100 125 150 175 200 s o TIME (SECONDS) N \\ x. 1 Figure 15.4-67 Transient Response to Steam Line Break at Exit of Steam Generator with ) Safety injection and Offsite Power (case b), Unit I L 4' 4 A-71 g g <," ', 3 't = g il ,, e ~ i - --- -- L

(.C:. ci.- Etc. d e gE 93 a' 4ce. O E8b t::. 22. 2EC. 7 s PRI$$URIZIR [MPTits AT 15 SIC 2::: \\ 8- "w Etc. EE h 10E0 at 500.

c. !

a-Y .2' b,8 .t5-gg .1 - w= n

1.

• INITIAL STEAM FLOW 1$ 2900 LB/SEC FROM THE FAULT [D STEAM GENIRATOR MD 8700 LB/5EC (2900 PER $C) 3g

., f FROM THE IKTACT STIAM GENERATOR 5 Eg o C E gB INTACT $ TEAM LINERATORS w >- 44 Uh (AULTED STIAM GINERATOR ONLY P ] 7 2300 PPM B3RON REACHES CORI AT 99 5tt a I~ =w W$ 0 '. / ~ -- 2y j ~ .) / - 2,, c. EP. 102 150. Pec. 25c. Etc. 5E0. ar2. 452. Etc. T !*E I EEC l FIGURE 15.4-67 DIABLO CANY.ON TRANSIENT RESPONSE TO STEAM LINE BREAK DOWNSTREAM OF FLOW MEASURING DEVICE WITH SAFETY INJECTION AND WITHOUT OFFSITE POWER (CASE C) A-72

7l93-El \\ - 600 \\ gar i l l l l 1 I '\\ U U ~ 500 \\O5g " 8. 400 s. g, 3000 PRESSURI ZER EM piles A1 19 SEC

• ~

am 8 t\\2000 "w \\ gg \\ g = 1000 5W a: n. 0 _\\ 3.0 \\ intil AL STEAM Flow 15 18261 tB/SEC FROM 'T AULTED S1EAM GENER ATOR ( AND 2983 LBS/SEC F' ROM IN1AC1 STEAM 8:N ER A10RS) 2.5- \\ g3 8 's / ( E-s s m x g $ 2.0 - \\ <g \\ / \\ O, d-1.5 - N wrzu

• dx

/\\ \\ $N $$ / 8D$$ FAULT [h sSTEAM GENERATOR ONLY \\ 5 1.0 E '\\ / \\ 0.5 - /' \\ / ~ 0 l " ' ~~ ~ ~ ~"' 2$ s7 / 20.000 PPM BORON RE ACHES LOOPS AT '35 SEC - sG/ \\

- g,/

o MU I l l l l l l ,/'E" s -2.5 0 25 50 75 100 125 150 175 200 7 / TIME (SECONDS) i i

1. 1 I

/ </h ( $P.H, t > ht Figure 15.4-68 Transient Response to Steam Line Break at Exit of Steam Generator with Safety injection and Offsite Power (case b), Unit 2 A-73

EC2. i;; Ete. 'a E.* :., w I g* 4c0. d' Wsg u ECO. 222. 2500. PRE 55URII[P. EMPTIES AT 10 SEC i ~ ff2000. 25cc. w $EEIa w::: EE EC2. t. .5< l h .25 ' l CN .2-I l $8 .15 - g5 .s-wa E .e5 - c-a U INIT!AL STEAM FLOW 15 9500 LB/SEC FROM THE 2.5 FAULT [0 STEAM GENERATOR AND 2900 L8/5EC (M6 PER SC)

g 3

-IROM THE INTACT STEAM GENERATORS 2. EU n.s - WE I "y i. g INTACT STEAM GENERATORS .5< FAULTED STEAM GENERATOR ONLY C 3, 9 E 2300 PPM BORON REACHES CORE AT 98 SEC _** 1 C. D. h2 ^ 0 $E r -1 -t e. 52. lec. 15e. 220. 25c. 520. 550. 422. TIME ISECl FIGURE 15.4-68 DIABLO CANYON TRANSIENT RESPONSE TO STEAM LINE BREAK AT EXIT OF STEAM GENERATOR WITH WATER INJECTION AND WITHOUT OFFSITE POWER (CASE D) A-74

l 7193-05 p 600 1 I I I I I I \\ g e_. 500 r- \\ !!!== ~ i N 8 N 300 \\yQ PRES $URIZER EMFilES A1 16 $lC \\= t 2000 o '9 55 t g N1000 <wEE \\ 'o 3.0\\ INiilAL F LOW 15 3092 LBS/$EC FROM FAUL1ED STE AM GENERATOR (AND 9276 LB/SEE FROM THE ilN1ACT STEAM GENERATORS) 2.5 -N \\ C Id W l5 2 2.0 g s S* uw "o $ $ \\ g,_ e - i.5 1 E5 "e \\ wg {g FAULhED STEAM GENERA 10R DNLY gU UN \\ g,g' u-

• g N'

U \\ 'n O.5 ~ / ~ 'j y=== = ~ ~ ~=== ~ / 7 2.5 / p; 20.000 PPM BORON RE ACMES LOOPS AT 33 SEC -c E-0 / / O5 E I I I I I -2.5 C 25 50 75 100 125 150 175 200 ~ TIME (SECONDS) Figure 15.4-69 Transient Response to Steam Line Break Downstream of Flow Measuring Nozzle With Safety injection and Without Offsite Power (case c), Unit i A-75 .\\ c.

7193-60 6 -C I I l l l l l \\ ..j 500 t =gs o* 400 3000 s-PRESSURIZER EMPTIES AT 16 SEC N ES am i ' 2000 xo, EE g g 1000 0U a: n. 0 3.0 INil AL FLOW A13092 LBS/$EC FROM F AULTED STEAM GENERATOR ( AND 9276 LB/$EC FROM s 2.5 - INTACT GENERATOR $) ii W o z ~ lg E 2.0 EE w" If u. 0 % "i 2 -5 2 uu.l.5 xy,o E5 $. $ o n. w o - m g 1.0 - FAULTED STEAM GENERATOR ONLY 0.5 ,' 0 2 2.5 4% 20,000 PPM BORON RE ACHE 5 LOOPS AT 43 SEC pc o /EU / em i I I I I I I / s. -2.5 ~ / 0 25 50 75 100 125 150 175 200 TIME (SECONDS) / l e -l Figure 15.4-70 Transient Response to Steam Line Break Downstream of Flow Measuring Nozzle With Safety injection and Without Offsite Power (case c), Unit 2 { A-76 h%' gf lu

7193-69 w 3000 \\ g I I I I I I I i se ao PRESSURIZER EMPTIES Al 19 SEC / R

• 2 2000 E-wwE

" E

• 1000

/ u\\ / ms o E\\ 600 .52 5eg\\500 / $ 5 E-1 \\ / ,,_ iloo =* g \\ 300 4 / 3.0 LOS/SEC FROM INITIAL STEAM FLOW 15'11253 ' F AULTED STEAM GENER ATOR ( AMD 2984 LBS/SEC - - FROM INTACT STEAp' GENERATORS)

2. 5

/ i2 0 n l h 2.0' / Es w c'w a g / +U-1.5 h $ N Lt. zg / o W5 wt g *o,..- w g*o _e i,

• E FAULTED STEAM GENERATORS ONLY i

0.5 / \\ / \\ / / / ne r===~w----- '2'

2. 5 20.000 PPM BORON REACHES LOOPS A135 SEC V<3 0

=

i PO / 1m i I I I I i 1 s -2.5 ~ / 0 25 50 75 100 125 150 175 200 f TINE (SECONDS) Figure 15.4-71 Transient Response to Steam Line Break at Exit of Steam Generator With Safety injection and Without Offsite Power (case d), Unit I g(T C A-77 l i

7193-6E \\ 3000 \\, et i I I I I I I PRESSURIZER EWilE$ AT 19 $EC '\\ N o E = 2000 g-s 'g EE g.y E 1000 =1 " \\ o mN 3\\ 600 $hL\\500 kl n \\ ~ o \\ u 300 3.0 , INITI AL STEAM FLOW 15 ll261/ LBS/SEC FROM \\ FAULTED STEAM GENERATOR ( AND 2963 LB/SEC FROM INT ACT STE AM GENERATOR $) 2.5 \\ j Is S / l;- = x s!w=" 2.0 N c em =K / ~_, e / O 5W %w I.5

• d x 25

/ \\ S M y i,o s / FAULTED'5 TEAM GENERATOR ONLY g s

0. 5,

/ / / s. 7, 7 2.5 20.000 PPM BORON REACHES LOOPS AT 37 $EC g f}<1 s /sE f p/ g,,,3 I I I I I I i ,/ 0 25 50 75 100 125 150 175 200 / TIME (SECONDS) / / R,u ~ \\e gt Figure 15.4-72 Transient Response to Steam Line Break at Exit of Steam Generator With Safety injection and Without Offsite Power (case d), Unit 2 A-78

7193 59 80

• 0 i

i i 20.000 / CASE A / p /.,/ 10,000 0 /,/ a S W.000 \\ / 5 40,000 CASE B / N i / 30,000 / oU l E 20,000 k / g y j- \\ / t 10,000 s w \\ g /\\ d \\ 10,000 o W \\ C'5C C E 5,000 \\ S / 5 0 / \\ \\ 10,000' / CASE D 5 \\ ~ 0 O 50 100 \\l50 200 [ TIME (SECONDS) \\ ,/ / / A [ h~e j' \\ () h (x

• g

/ ,M,)J q Figure 15.4 73 Integrated Flowrate of Borsted Water vs. Time, (Unit 1) A-79 i

- 20, - a b [ CASE A 10 : h~ I v M O E O 200 400 30 e E e 6 20 - CASE B W e is 10 - 5 u 0 0 100 200 300 400 20 n 9 6 / 5 10 i CASE C s w ~ v 0 0 200 400 20 : .a E 6 g CASE D ce 10-2 g 8 0 0 100 200 300 400 TIME (SEC) FIGURE 15.4-73 DIABLO CANYON CORE BORON YS. TIME A-80 m-

l 7893-60 40,000 \\ i i i \\ 30,000 l \\ CASE A \\ + 20 g ,000 0.000 ' / 0 E S / 50,000 \\ j 40,000 x \\ CASE 8 \\ eg 30,000 \\ E" 20,000 'N \\ g \\ 10,000 \\ m \\ a: O Ed o y 10,000 \\ g y g S,000 casg c r \\ o \\ \\ 10,000, \\ CASE D 5,000 \\j / l f' 0 / 0 50 lo0 150\\ 200 / TINE (SECONDS) / \\ ) Av r l gcb / Figure 15,4 74 Integrated Flowrate of Borated Water vs. Time, (Unit 2) A-81 \\.

i, WESTINGHOUSE PROPRIETARY CLASS'3 APPENDIX B TECHNICAL SPECIFICATION REVISIONS FOR BIT ELIMINATION O ,Y B-I

_ _ _ _ - _ _ _ _ = _ _ _ _ _ EMERGENCY CORE COOLING SYSTEMS BASES ECCS SUBSYSTEMS { Continued) TherequirementtomaintaintheRHRSuctionValves8701$tnd8702inthe locked closed condition in MODES 1, 2 and 3 provides assurance that a fire could not cause inadvertent opening of these valves when the RCS is pressur-ized to near operating pressure. These valves are not part of an ECCS subsyster The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all centrifugal charging pumps and Safety Injection pumps except the required OPERABLE charging pump to be inoperable below 323*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The Surveillance Requirements provided to ensure OPERABILITY of each component ensurws that, at a sinimum, the assumptions' used in the safety analyses .are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) present total pump flow from exceeding runout conditions when the system is in its einimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed -in the ECC5-LOCA analyses. 3/4.5.4 BCRON INJECTION SYSTEM jy The OPERABILITY of the Boron Injection System as part of the ECC5 ensures that ufficient negative reactivity is injected into the core-to counteract any positive increase in reactivity caused by RCS cooldown.,AtCS cooldown can be s caused byKadvertent depressurization, a loss-of-cociant accident or a steam line rupture. The limits oni n tion ensure that the%jection tank minimum dontained volume and boron concentra-sumptions used 4fi the steam line break analysis are met. The contained water volung limit includes an allowance for water not usable because of tenk tlischarge h'ie)tiration er other physical characteristics. The OPERABILITY offe redundant heat tracing channels associated with the boroninjectionsyste ensure that the solubility of the boron solution will be maintained abov he solubility limli ' f 135'F at 21.000 ppm boren. O f a DIABLO CANYON - UNITS 1 & 2 8 3/4 5-2 B-2

l Dele EMERGENCY CORE COOLING SYSTEMS 3/4. 5. 4 BORON INJECTION SYSTEM / BORON INJECTION TANK a LI JTING CONDITION FOR OPERATION 4 3.5.4.1 The boron injection tank shall be OPERABLE with: minimum contained borated water volume of 90' gallons of berated a.. w er, b. A be n concentration of between 20,000 a 22,500 ppm, and c. A minimkn solution temperature of 145'. APPLICABILITY: MDD 1, 2 and 3. ACTION: With the boron injection tynk inoperab, restore the tank to OPERABLE status within I hour er be in HOT STANDBY and berated to a SHUTDOWN MARGIN ec;uivalent s to 1% Ak/L at 200'F within thq next 4 hours; restore the tank to OPERABLE f status within the next 7 days b be in HOT SHUTDOWN within the next 12 hours. i SURVEILLANCE REQUIREMEh75 4.5.4.1 The boron i tion, tank shall be monstrated OPERABLE by: Verifyipg the contained borated wate volume through a recirculation a. flow test at least once per 7 days, b. Ver) ying the boron concentration of the ater in the tank at least once per 7 days, and / c. erifying the water temperature at least once r 24 hours. 6 l DIABLO CANYON - UNITS 1 & 2 3/4 5-9 B-3 L____--____________-___-_______________________-_____

EMERGENCY CORE COOLING SYSTEMS HEAT TRACING 1 / LIMITING CONDITION FOR OPERATION ./ y j s 3.5.4.2 At least two independent channels of heat tracing shall be OPERABLE for the boron injection tank and for the heat traced portions of the associated flow paths / APPLICABILITY: MODES 1, 2 and 3. ACTION: With only one channel of heat tracing on either the. oron injection tank or on the heat traced portion of an associated flow path'0PERABLE, operation may continue for up to 30 days provided the tank and Tiow path temperatures are s verified to be greatersthan or' equal to 145'F at least once per 8 hours; otherwise, be in HDT STAf(DBY within 6 hours an'd in at least HOT SHUTDOWN within the following 6 hob s.- SURVEILLANCE REQUIREMENTS / N 4.5.4.2 Each heat tracing cinannel for th boron injection tank and associated flowpathshallbeder.onspatedOPERABLE: At 1 east once/per 31 days by snergi 'ng each beat tracing channei s. and b. At least/ence per'24 hours by verifying he tank and flow path tem-peratu s to be greater than or equal to 145'F. The tank temperature shall e determined by measurement. The f) path temperature shall be termined by either measurement or reci ulation flow until as lishment of equilibrium temperatures wit in the tank. p DIABLO CANYON - UNITS I & 2 3/4 5-10 B-4 _ - _ _ - _ _ _ _ _ _ _ - - -. _ _ _ _ _ _ _ _ _ _ _ _ _.}}