ML20067C266

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Proposed Tech Specs Re Setpoint Tolerances for Main Steam Code Safety Valves
ML20067C266
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 02/04/1991
From:
SOUTH CAROLINA ELECTRIC & GAS CO.
To:
Shared Package
ML20067C264 List:
References
NUDOCS 9102120007
Download: ML20067C266 (37)
v • d • e


Text

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ATTACHMENT 1 MARKED-UP TECHNICAL SPECIFICATION PAGES .

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Pogg ,$pecification Chance Description 1 3/4 7-1 3.7.1.1. Turbine Cycle. Delete Action b. Rename Safety Valves Action c as Action b.

3/4 7-2 5.7.1.1 Turbine Cycle. Delete Table 3.7-2.

Safety Valves 3/4 7-3 3.7.1.1. Turbine Cycle, Reformat Table 3.4-3 per Safety Valves insert 1. Also renumber Table '

3.7-3 as 3.7-2 and move to page 3/4 7-2 0 3/4 7-1 3/4.7.1.1 Bases. Turbine Delete references to two loop Cycle operations.

B 3/4 7-2 3/4.7.1.1 Bases. Turbine Delete references to two loop Cycle operations.

l 9102120007 910204

, PDR ADOCK 05000395 P PDR l_

_ - - . . . ,,__.... u-. _ . _ - . , - _ _ , - . , - _ _ - ,

t 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION _

3.7.1.1 All main staae line code safety valves associated with each steam generator shall be OPERABLE with Ilf t settings as specified in Table 4r7-Or J . 7 - 2.

APPLICABILITY _: H0 DES 1, 2 and 3.

ACTION:

a. With 3 reactor coolant loops and associated steam generators in operation and with one or more main steam line code safety valves inoperable, operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Range Neutron Flux High Trip Setpoint is reduced per Table 3.7-1; otherwise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

1

_'bNWith 2 reactor coolant loops and associated steam gensratott in 7

operation-tod with one or more main steam line code-saf4t~y valves associated wlDrasop(rating loop inopetable,Tperation in MODES 1, 2 and 3 may proceed provTdedr-.that-within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoper-able valve is restogd to-0PERABLT'statut or the Power Range Neutron Flux High . Trip-$st~ point is reduced per Tab 1MMLothemise, be in apeast~ BOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in C0tD4HLIIDOWN N

, M i [in the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.  %

% The provisions of Specification 3.0.4 are not appitcable,

b. .

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.

SL"MER

  • UNIT 1 3/4 7-1

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l TABLE 3.7-1_

l MAX 101 ALLOWABLE POWER RANGE flEUTRON FLUX HIGH KTMIT STLAM LINE TIT [TV VAlyES DURIM 3 LOOP OPERATIOT Maximus Allowable Power Range Maximus Number of inoperable Neutron Flux High Setpoint Safety Valves on Any gjecent of RATED THERMAL POWER 1 3

AeratiaSteamGenerator 67 1

65 2

43 3

ISt%fff ,

_N TABLE 3 7-2 .

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MAXIMUM ALLOWASLE POVER f!ANGE NEUTPON FLUX HIGH

~ Ih0PERABLE 51EAM LINE SAFETY / VALVE 5 DURIE 2 LOOP __0P y

Maximus--Allowable Power Range Maximue N wber of Inop r.4ble Heutron Flux Hign Setpoint Safety Valves on Any N Operatina Storm Generator

  • iPe'rcent of RATED THERMAL POWER) n.

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x \

3

-[

,/ team "3t'leasttwosafetyvalvesshallbeOPERABLEonthenon-oparatin

- ,/ generator."'These values lef t blank pending NRC approval of two-loop operation.

N 4

I SUMMER UNIT 1 3/4 7-2

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4 INSERT 1 TABtf 3,7-2 STEAM LINE SAFETY VALVES PER LOOP S/G A S/G 8 S/G C Lift Setting

  • Orifice Size XVS-2806A XVS-2806f XVS-2806K 1176 psig 11% 4.515 In dia/16 sq in XVS-280661 XVS-2806G XVS-2806L 1190 psig 13% 4.515 In dia/16 sq in XVS-2806C XVS-2806H XVS-2806M 1205 psig 13% 4.515 In dia/16 sq in XVS-28060 XVS-28061 XVS-2806N 1220 psig 13% 4.515 in dia/16 sq in XVS-2806E XVS-2806J XVS-2806P 1235 psig 13% 4.515 In dia/16 sq in
  • The lift Setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure, s

kf Pt H E WI&J I/VS tTAi l pwp pp,c p f3;; ,=/& 7-2

.I } '-i j' TABLE 3.7-3 E '

f STEAM LINE SAFETY VALVES PER LOOP E

, VALVE MUMBER LIFT SETTING # '%)" ORIFICP IZE C

'5/G A

-.i  :

Xv5-2806 A- 1176 psig 4.515 In dia/16 sq in l XVS-280s B 1190 psig 4.515 In dia/16 sq in l n5-2806 C 1205 psig 4.515 In dia/16 sq in Xv5-2806 D' 1220 psig 4.515 In dia/16 c:a in AVS-2806 E 1235 psig 4.51L In dia/16 sq in 5/G B XVS-2806 F T176 psJr. 4.515 In dia/16 sq in XVs-7806 G 11%gir; 4.515 In dia/16 sq in ,

XVS-2806 H 1216 psig 4.515 In dia/16 sq in  :

" XV5-2806 1 1220 psig 4.515 In dia/16 sq in

  • XVS-2806 J 1235 psig 4.515 In dia/16 sq in 7

" 5/G C XVS-2806 K 1176 psig 4.515 In dia/16 sq in XVS-2806 L' 1190 psig 4.515 In dia/16 sq in XVS-2806 M 1205 psig 4.'515 In dia/16 sq in  ;

XVS-2806 N 1220 psig 4.515sJn dia/15 sq in i XV5-2806 P 1235 psig 4.515 Insdia/16 sq in i

  • Jh6 Lif t Setting pressure shall correspond to amblect conditions of the valve at nominal fo >erating tcoperature ' and f,ressure.

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4 3/4.7 PLANT SYSTEMS BASES -

3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary sy. stem pressura will be limited to within 110% (1305 psig) of its design pressure of 1185 piig during the most severe anticipated system operational transient. TF4 waximus rcHeving capacity is assor,ieted with a l turbine trip from 100% RA'lED THERMAL POWER coincident with an assumed loss of condenser heat sink (f.e. , no steam bypass Lv the candecce?). l The specified valve lift settings and rslieving capacities are in accoro-ance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Editien. The total relieving capacity for all valves on all of the l steam lines is 13.76 x IOS lbs/hr which is 110 percent of the total secondary I steam flow of 12.2 x 10e lbs/hr at 100% RATED THERMAL POWER.  :. minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity *is available for thc allowable THERMAL POWER restriction in Table j? 7-l '3r7-2; STARTUP and/or POWER OrERATION is allowable with safety valves inoperable within the limitations of ine ACTION requirements on the basis of the reduction l in secondary system steam flow and THERMAL POWER required by the reduced  !

reactor trip settings of the Power Range Nostron flux channels. The reactor J trip setpoint reductions are derived on the, i'>llowing bases: J For 3 loop operation 3p , {_X) - (Y)(V) 109 X X (OhL p oportti PMU).~

}Pefr - -

p Where:

$P = Reduced reactor trip setpoint in percent of RATED THERMAL POWER V = Maximum number of inoperable safety valves par steam line

-- U -- --= Maxtmum-nu.nber-of-htoperat4e-s a f ety-valves-per-ope rat 4ng--

+tean 14ne.

SUMMER - UNIT 1 3 3/4 7-1 l

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PLANT SYSTEMS BASES SAFETY-VALVES (Continued) 109 m - Power Range Neutron Flux-High Trip Setpoint for 3 loop operation.

".ukte p;rcer.t ef RATE 9-THEAMM.-PCMER-permiss%)e-by-P-8

---Setpoint-for-2-loop-operationr-This-value-lef t-blank-pending-

-NRC-epproval-o f-2-loop-ope ra ti on e X = Total relieving _ capacity of all safety valves per steam line in Ibs/ hour.

o Y = Maximum relieving capacity of any one safety valve in 1bs/ hour.

3/4.7.1.2 EMERGENCY FEEDWATER SYSTEM The OPERABILITY of the emergency feedwater system ensures that the

(_~ Reactor Coolant System can be cooled down to less than 350'F from normal-

i. operating _ conditions in tha event of a total loss of off-site power.

' Tne emergency feedwater .ystem is capable of delivering a total feedwater flow of 380 gpa at s' pressure of 1211 psig to the entrance of at least two l steam generators while allowing for (1) any spillape through the design L

worst-case break of the emergency feedwater line, (2) the design worst-case

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ningle failure, and-(3) recirculation flow.. Thislcepacity is sufficient to ensure that idequate feedwater flow is available to remove decay heat and reduce the Neactor Coolant: System temperature to less than 350*F at'which L- point the Residual Heat Removal System may be placed into operation.

L 3/4.7.1.3 CONDENSATE STORAGE TANK p

2The OPERABILITY of the condensate storage tank with the minimum water l.- -

volume ensures that sufficiant' water is available to-maintain the RCS at HOT STAND 8Y conditions for 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> with steam discharge.to the atmosphere con-current with. total-_ loss of_offsite power. The contained water volume limit includes an allowance for water.not usable because of tank discharge line

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location or other physical characteristics.

3/4.7.1.4- ACTIVITY The ? imitations on secondary system specific activity ensure that the resultant offsite radiation dose will be _ limited to a small fraction of 10 CFR Part 100 lirits in the event of a steam line rupture. -This-dose also includes .

the effects of a coincident 1.0 GPM primary to secondary tube leak in the .

steam generator of the affected staam line. These values are consisteret with-the assumptioni used in the accident analyses.

l SUP91ER - UNIT 1 9 3/4 7-2 y

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ATTACHMENT 2 DESCRIPTION OF AMEN 0 MENT REQUEST SAFETY EVALUATION I

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$ Attachment 2 to Document Control Desk letter

. TSP 880018-0 Page 1 of 3 DESCRIPTION Of AMENDMENT REQUEST This arnendment request involves Technical Specification (T.S.) 3/4.7.1.1

" Turbine Cycle - Safety Valves" and addresses two separate changes. The first change is strictly administrative in nature, and the second change is a request to modify the acceptable setpoint tolerance assouiated with the Main Steani Safety Valves (MSSVs).

Presently, T.S. 3/4.7.1.1 contains provisions which were included for the NRC's eventuai approval of two-loop power operations. These provisions consisted of Action Statement b. which specifically applies to two-loop operation, and table 3.7-2 which prescribes the maximum power allowed during two-loop or;eration based on the number of inoperable MSSVs. SCE&G is requesting the removal of these provisions based on the fact that it appears highly unlikely that two-loop operation will be approved. SCE&G is also concerned, from a human factors perspective, that having irrelevant matter in the specification could be detrimental to its application.

The second item to be addressed involves the setpoint tolerance for the MSSVs. The current Limiting Condition for Operation (LC01 equires that the MSSVs be operable with lift settings as specified in Table . 7-3. It is important to note that there are five MSSVs per steam line, and that the setpoint of each of the five valves increases sequentially in incre.aents of essentially fifteen psig (i.e., the lowest setpoint valve on each steam line is 1176 psig, the next is 1190 psig, the next is 1205, and so on). Table 3.7-3 currently imposes a 1% telerance on all of the MSSV lift setpoints.

SCE&G is requesting that the lift setpoint tolerance be increased from 11% to 1 3% for the four highest set MSSVs per steam line while maintaining 1 %

1 as the setpoint tolerance on the lowest set MSSV per stear,i line.

The operability of the MSSVs ensures that the maximum pressure experienced by the secondary system will be limited to 110%-(1305 psig) of design pressure (1185 psig) during the most severe anticipated transient. T.S. 3/4.7.1.1 requires that the MSSVs be tested and verified operable in accordance with

-Section XI of the ASME Boiler & Pressure Vessel (B&PV) Code. The code does not contain a setpoint tolerance; therefore, the 11% setpoint tolerance prescribed in T.S. 3.7.1.1 is applied as an acceptance criteria. SCE&G proposes to increase the setpoint tolerance to 13% based on the advancement in technology which can more accurately # termine the lift setpoint and the inability to make the corresponding fine adjustments to the MSSVs. Also, an evaluation of a 3% tolerance shows that the related effects of a larger setpoint tolerance yields no safety concerns and does not prevent the MSSVs from performing their design function.

. . Attachment 2 to Document Control Desk Letter TSP 880018-0

, Page 2 of 3 SAFETY EVALUATION:

An evaluation was performed to ensure that increasing the MSSV setpoint tolerance f rom i is to 13% did not compromise safety. The impact of the toleratice change was assessed with respect to the following areas: The ASME B&PV Code, Westinghouse's Safety Analyses, Gilbert Commonwealth's Safety Analyses, and the Technical Specification Margin of Safety.

1 The MSSVs are in compliance with the ASME B&PV Code Section 111 (1971 edition, Winter 1972 addendum) which provides requirements for the design of the MSSVs. No requirements exist in Section 111 regarding the tolerance on the lift pressure setpoint. However, there is a requirement in Section 111 that the valve itself be designed to have a popping point tolerance of i 1%

(i.e., the repeatability of the valve is within i 1%). The Inservice Inspection (ISI) required by T.S. 4.7.1.1 is in compliance with Section XI of the ASME B&PV Code ('77 ed., S'78 add.); also, this version of Section XI 3 does not specify a setpoint tolerance requirement. However, the 1989 Edition J of Section XI currently refers to ISI guidance which states that safety valves must not exceed their stamped set pressure by 3% or greater.

, Therefore, a MSSV setpoint tolerance of 1 3% does not contradict the B&PV

< Code currently committed to and is consistent with the most recent ISI guidance provided by the Code.

Westinghouse performed a safety evaluation to address increasing the setpoint tolerance of all the MSSVs from i 1% to i 3% with respect to its effects on the Reactor Coolant System (RCS) and the LOCA and non-LOCA licensing basis events. This evaluation included an analysis of each event that is discussed in Virgil C. Summer Nuclear Station (VCSNS) Reload Transition Safety Report (RTSR), an analysis of each FSAR LOCA related analysis. and an analysis of VCSNS's Steam Generator Tube Rupture Analysis. The results of the Westinghouse evaluation concluded that all licer..iing basis criteria continue to be met, and the conclusions in the RTSR remain valid.

Westinghouse also published a letter (CGE-90-1157) which, based on an examination of the VCSNS icensing Basis Analyses, revealed that the maximum relief capacity required by the MSSVs to satisfy the most severe anticipated transient was 82.:7 of rated steam flow. A subsequent letter (CGE-90-1160) verified that 82.3% total relief capacity can be substituted in lieu of the 110% value provided in section V-2 of the Westinghouse Steam Systems Design Manual. Provided with this information, Gilbert / Commonwealth (G/C)--the Architect Engineer for VCSNS Steam Sysrems--performed a calculation (G/C calculation DC-501-0428-11, Rev. 0) and verified that only four of the five MSSVs are required to meet the licensing basis events. Based on the results of thcir ca tulations, G/C performed an analysis of the increase in the MSSV's setpoint tolerance from 1% to 13% and concluded that the increased tolerance does not affect the MSSV's ability to perform their design function.- However, it was discovered that the setpoint of the lowest set MSSV is used in the Emergency feedwater (EFW) System capacity calculations.

A change to 3% for this MSSV could affect the EFW system's capability to meet the Westinghouse criterion of a maximum 65%/35% split of EFW flow to any two steam generators. Therefore, the change in setpoint tolerance of the lowest set M55V is not included in this amendment request. ,

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Attachment 2 to Document Control Desk Letter

, TSP 880018-0 Page 3 of 3 Finally the Technici Specification margin of safety was evaluated. The purpose for the MSSV's as described in the Bases is to limit the most severe anticipated transient to 110% of design pressure (1305 psig) and to maintain the lift settings and capacities consistent with Section III of the ASME B&PV Code, 1971 edition. Therefore, the T.S. margin of safety is the margin between 1305 psig and the pressure at which ultimate failure of the secondary pressure boundary occurs. As previously stated, an examination of the licensing bases for VCSNS has shown that changing the MSSV's setpoint tolerance from 11% to i % does not cause an increase in the maximum upset pressure and is consistent with Section 111 of the ASME B&PV Code, 1971 edition. Thus, the T.S. margin of safety is not affected by this change.

Based on an indepth review of the evaluations described above, SCE" has concluded that the amendment request continues to meet the requirements of the ASME B&PV Code and involves no significant increase in safety consequences.

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ATTACilMENT 3 DESCRIPTION Of AMENDMENT REQUEST N0 SIGNIFICANT llAZARDS EVALUATION 1

Attachment 3 to Document Control Desk Letter TSP 880018-0 Page 1 of 2 l l

DESCRIPTION OF AMENDMENT REQUEST:

This amendment request involves Technical Specification (T.S.) 3/4.7.1.1

" Turbine Cycle - Safety Valves" and addresses two separate changes. The first change is strictly administrative in nature, and the second change is a request to modify the acceptable setpoint tolerance associated with the : lain Steam Safety Valves (MSSVs).

Presently, T.S. 3/4.7.1.1 contains provisions which were included for the NRC's eventual approval of two-loop power operations. These provisions consisted of Action Statement b. which speci'ically applies to two-loop operation, and table 3.7-2 which prescribes ;he maximum power allowed during two-loop operation based on the number of iroperable MSSVs. SCE&G is requesting the removal of these provisions based on the fact that it appears highly unlikely that two-loop operation v/'ll be approved. SCE&G is also concerned, from a human factors perspeci1ve that having irrelevant matter in the specification could be detrimental to its application.

The second item to be addressed involves the setpoint tolerance for the MSSVs. The current Limiting Condition For 2peration (LCO) requires that the MSSVs be operable with lift settinos as specified in Table 3.7-3. It is important to note <ot 'here are <1 F .Vs per sieam line, and that the setpoint of each or 'ive valves nereases tequentially in increments of essentially fifteen psia (i.e., the lowest setroint valve on each steam line is 1176 psig, the nex'. is 1190 psig, the next is 1205, and so on). Table 3.7-3 currently imposes o 11% tolerance on all of the MSSV lif t setpoints.

SCE&G is requesting that the lift setpoint tolerance be increased from 11% to 13% for the four higb9st set MSSVs per steam line while maintaining 11% as the setpoint tolerance on the lowest set MSSV per steam line.

The operability of the MSCVs ensures that the maximum pressure experienced by the secondary system will be limited to 110% (1305 psig) of design pressure (1185 psig) during the most severe anticipated transient. T.S. 3/4.7.1.1 require 3 that the MSSVs be tested and verified operable in accordance with Section XI of the ASME Boller & Pressure Vessel (B&PV) Code. The code does not contain a setpoint tolerance; therefore, the 11% setpoint tolerance prescribed in T.S. 3.7.1.1 is applied as an acceptance criteria. SCE&G proposes to increase the setpoint tolerance to 13% based on the advancement in technology which ,2n more accurately determine the lift setpoint and the inability to make the corresponding fine adjustments to the MSSVs. Also, an evaluation of a 13% tolerance shows that the related effects of a larger setpoint tolerance yields no safety concerns and does not prevent the MSSVs fro" am orming their design function.

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Attachment 3 to Document Control Desk Letter TSP 880018-0 Page 2 of 2 No Significant Hazards Determination:

This amendment request has been reviewed with respect to Title 10 of the Code of Federal Regulations (10CFR) part 50.92 and found to contain no significant hazards considerations for the following reasons:

1) The amendment request does not involve a significant increase in the probability or consequences of an accident previously evaluated. The effects of the requested change was examined with respect to each everit described in the RTSR (non-LOCA events), the small and large break LOCA accidents, end the Steam Generator Tube Rupture Event. The examination revealed that the conclusions reached for all events described in the RTSR remained valid and the results of the FSAR accident analyses were not impacted.
2) The amendment request does not create the possibility of a new or different kind of accident from any accident previously evaluated. The requested change does not represent a design change in that all design limits are maintained and the physical design of all systems are unaffected. Therefore, the potential for malfunction or failure of any component or system as a result of the requested change remains unaffected.
3) The amendment request does not involve a significant reduction in a margin of safety. The rtquested change does not affect the minimum or maximum pressures experienced by the main steam system during any licensing basis event and remains consistent with the margin of safety as described in the bases of the Technical Specifications.

Again, for the reasons listed above and supported by the attached safety evaluation (Attachment 4), SCE&G has determined that the requested amendment to T.S. 3/4.7.1.1 has no significant hazards considerations.

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ATTACHMENT 4 SUPPORTING DOCUMEKTS 10CFRSO.59 evaluation supporting an increased Main Steam Safety Valve lift pressure setpoint tolerance of 13%.

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Page i c(3 Revision 0-0 VIRGIL C. SUMMER NUCLEAR STATION 10CFR50.59 SAFETY EVALUATION WORRSHEET Check Applicable Yes( } and No( ) Indications pantNToccuMINT Does this evaluation change the Final Safety Analysis Report or Fire Protection Evaluation Report?

TECH SPEC. REFERENCE *l Section Page FSAR/FPER REFERENCE Yes( )

No W Chapter 3 . 7. /. / M P3 Section Dage Tnbs1 7,74 3,7 3 "

15 a chang e in Tech gg

[3 g 7.[ Specific ioriinvolved)

Not accressed in 2 Tech 50ecs ( } yes% No() Not addressed in e n pgggjppgg ( )

[

UNREVIEWED SAFETY QUES TION DETERMINATION:

Answer the seven Questions on pages 2 and 3. Provide

/ justifying the decision for the "yes" or "no" answers. specific reason Nuc uc Reviewer CIiie i

NOTE:

o PSRC/NSRC Review Restatement of the ouestion in a negativ sense or makmg a v: tole statement of conclusion is not sufficient and shall be avoider recogn>2ed, however, that for certain very simple activitie' . a '

f statement of the conclusion with identification of refcre ,ces consulted to support the conclusion will be acequate.

Reovest and Receive Nuclear Regulatory Commissinn Authorizat!on for Change Prior Complete the items below after the questions on pages 2 ard 3 nave been accressed.

To im plementation Of the SuDject Change v

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Authorization Denied

  • 3]1 Any Answer Yes (f} l All Answers No ( }

v v

Abort Authorizatson initiat6 -

[ The Change ~

Received The Change

'If answer (C is "yes" but answers (2) and (3) are no,

then Ine change is reportable under 10CFR50.59b and a A. 4 AU% L t.ead Engineer / Preparer /

%hi

') ate description of the change will be included in the Ant'ual Report if answer f 2)is "yes" the,*i 10CFR50.59 is not t','

l- .'7'*' "" ~

II*D app 1 stable. Proceed to 10CFR50.90 "# 'D'"O '"I 8 '* "

kes' W b /n,,

Approval 5ignature i <l3t)l4 xe

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EN0INEER8 Burial 239-02-7_831_

Engineer 6.4 M w %

TECHNICAL WORK RECORD Date 1/23/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 1 of 22_

South Carolina Electric & Gas Co. Wishes to submit a Tech.

Spec. change to increase the set point tolerances of the four (4) highest set Main Steam Safety Valves (MSSV) on each loop from +/-

1% to +/- 3%. The lowest set MSSV on each loop will remain the same (+/- 1%). The following paragraphs show that the Tech. Spec.

margin of safety for the MSSV set point tolerances has not been reduced.

PLANT TECHNICAL BPECIFICATIONS - M88V's Piring normal operation, Tech. Spec. 3.7.1.1 (Table 3.7-3) curror.cly requires that the MSSV's be set as follon.:

MAIN STEAM SAFETY VALVE 8 M88V SET POINT MAIN STEAM. MAIN STEAM MAIN STEAM +/_ ig LOOP A LOOP D LOOP C XVS-2806 A XVS-2806 F XVS-2806 K 1176 PSIG XVS-2806 B XVS-2806 G XVS-2806 L 1190 PSIG

-XVS-2806 C XVS-2',06 H XVS-2806 M 1205 PSIG XVS-2806 D XVS-2C)6 I XVS-2806 N 1220 PSIG XVS-2806 E XVS-2806 J XVS-2806 P 1235 PSIG A visit to the Bases for this particular Tech. Spec. indicates that the MSSV's are required to operate to prevent the secondary system pressure from exceeding 110% (1305 psig) of its design pressure (1185 psig) . The Bases for this Tech. Spec. state that the valve lif t settings and relieving capacities are in accordance with ASME B&PV Code Section III requirements. The ASME code requires that anticipated transient pressures cannot exceed 110% of design pressure. Since design pressure is 1185 psig (1200 psia) then 11( %

of design pressure is 1305 psig (1320 -psia) . Hence the Tech. Spec, secondary system limit is the same as that required by the '4ME Code. The ASME Code margin of safety consists of the area bet .n 110% of design pressure and the pressure which causes ultimate failure of the pressure boundary. In this case, the ASME Code marg 3h of safety and thus the real Tech. Spec. margin of safety is the r.rea between 1305 psig and ultimate failure of the secondary system pressure boundary.

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B ENGINEERB Berial 239-02-7834 Engineerd,6 fr; TECHNICAL WORK RECORD Date 1/21/f_1 Project Title _MSSV SET _EoINT TOLERANCE INCREASE Tab Page 2 of._22

  • ASME B&PV CODE The AS!'E B&PV Code Section III_ ('71 ed., W' 72 add. ) under subsection NC-7411 -requires that the " total rated relieving .

capacity shall, ..., be sufficient to prevent a rise in pressure of

-more than 10 per cent above system design pressure ... under any pressure transients anticipated to arise"._ NC-7511 further requires-that at least one safety valve be set at the system design pressure. NC-7512 requires that -pressure drop including back pressure be considered in meeting the 110% of design - pressure i requirement. No mention .is made of any requirement to con' ider any' l_ safoty valve set _ point tolerance in the system design. dC-7614.3 requires that the safety valve itself shall have a popping-point tolerance of +/- 1%.- .However,1 Section XI ( '77 ed. , S ' 78 add. ' does not-specify a set point tolerance.

OM 1 specifies that the. valves when tested shall_not exceed the'r~ stamped set. pressure by 3%-or greater. The-corresponding ASP B&pV Code Edition & Addenda for Section III subsection NC-7000 do 1.ot address the SI; tolerance for valves with set pressures over 1000 psig.

WESTINGHOUSE SAFETY ANALYSE 8 i >

l. Current plant design-incorporates 5 MSSV's on each_ loop with

-a combined-name plato capacity'of.110%.of full rated flow lat.(100%

reactor _ pow.r. The set points on the LMSSV's ' for each loop are h staggered in banks.- Etch: bank consists of one valve'on each_ loop (three valves total, all- having- the same. set point) . Each bank-(five total) has a capacity.of_22% of full rated flow.

1 The Westinghouse safety analyses for anticipated transients is : <

bounded by. the " Loss of; Load / Turbine Trip ~ 9100% Power" ev2nt for o , over-pressurization - events. - ~This event . requires a capacity of-

! 82.3% of fullirated flow in order.to keep. maximum secondary system L pressure below 110% of design pressure. Therefore, the valves-which have a combined capacity of 82.3% are all that is required to meet the-lASME code.-requirements and -thus meet Tech. Spec.

requirements. - Since, four-banks of valves have a combired capacity of.88%,-tho'fifth bank of valvesEis.not required.

NOTE: . It should be noted that the difference between- the IWestinghoqse s.atelv e analysgs - reqqirJments - (82.3% full rated flowl. >

1'

]

l

_ _. _ _ _ . . . - . _ . _ _ ~ . - _ . _ _ _ _ _ . . _ . . _ _ _ - _ _ _ __.___.

e ENGINEERS Serial 239-02-7834

- Engineer d.4. A-s TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 3 of_22 l and the ASME / Tech. Spec. reauirements is desian marcin not Tech.

Spec. marain.

The following paragraphs address the specific safety analyses looked at by _ West: nghouse. It is a. synopsis of- Westinghouse

, Nuclear Safety Evaluations No.'s SECL 89-939' and 1140.

No technical basis has been changed.

Historically, the 1% -tolerance of the Pressurizer Safety Rolief Valve (PSRV) and the Main Steam Safety Valve (MSSV) set pcints has been negligible with respect t *'e safety analyses: and ,

thus, has not been_ accounted for. However, an increase In_the

. tolerance to +/- 3% is-considered to be sufficiently significant such that its impact on the safety analyses should be considered.

__ Modifying : either side of the tolerance band potentially affects the cofety analyses. .The PSRV's and MSSV'r provide l' _orutection_from over-pressurization of the primary,and.=ocondary ,

systems,: respectively. By increasing the positive ride of the tolerance band, the pressure at which the safety valve' potentially

'lif t and thus the potential maximum pressure attained .s increased.

By ' increasing-the negative side of.the tolerance band, the pressure at which the safety valves potentially lift is-decreased.

A Tech. Spec. change has previously been submitted - to - and approved by;the NRC to increase the PSRV set point tolerance.to +/-

3 % . :. As.a: result, this' evaluation' conservatively assumes thatLthe-valve -lift -set --points for'~ both the PSRV's and~ the MSSV's are

increased- to -+/- 3%. Furthermore, it is assumed that the; accumulation point for - the PSRV's - and the MSSV's? occurs ~t a a-pressure 3% above the actual valve lift set point. This is more conenrvative than tho'ASME code requirement which states.that the accumulation point occur within 3% above the nominal valve lif t set-

_ point-for the valve.

li *

! 'RON-LOCA-L Each non-LOCA licensing event is discossed below in the order in which it appears in the Reload Transition Safety Report (RTSR) for the Virgil C. Summer Nuclear Station'(VCNS).

n  :.-

= ENGINEERS Serial 239-02-7834-Engineerd.4.h sw -

TECHNICAL WORK RECORD Date 1/21/91 -;

Project Title _}i9SV SET POINT TOLERANCE INCREASE _ Tab- Page 4 o f_22

1. Uncontrolled Rod Cluster Control Assembly (RCCA) Bank Withdrawal from a Subcritical Condition (RTSR Section 15.2.1)

For this condition II event, rod withdrawal results in a rapid reactivity insertion and increase in core - power potentially leading to high local-fuel temperatures and heat fluxes and a reduction in the minimum DNBR. The transient is promptly terminated by a reactor trip on the Power Range High Neutron Flux - low set point. ,Due to the inherent thermal lag in the fue: pellet, heat transfer to the RCS is relatively slow and the minimum DNBR.is shown to remain above the limit value. No credit is taken for the MSSV's. Tms, the results of this analysis are unaf fected by increasing t.1e tolerance on the MSSV's to +/- 3% and the conclusions in tl'e-RTSR remain valid.

2. Uncontrolled Rod Cluster Control .% cMiy _ (RCCA) Dank Withdrawal at' POWER (RTSR Section 15.2.2)

.For this condition II event, various initial power levels and reactivity insertion rates for both minimum and maximum feedback _ assumptions are ._ analyzed. The resulting power excursion may;1end to high-local fuel' temperatures and heat fluxes and-a reduction in the minimum DNBR. Since this= event-is.a' limiting DNB-event and.not peak pressure limiting,.-the

-Pressurizer PORV's are conservatively assumed lto be operable.

Neither the ~~ primary nor- the - secondary systems- reach = ithe reduced safety set point during this _ event. Thus, the results of this analysis are unaffected by increasing the tolerance on thesMSSV's to +/- 3_% and the conclusions in the)RTSR remain-

, valid.

~3. 1 Rod ' Cluster- Control--Assembly Misoperation (RTSR ' Section 15;2.3)

. This condition II event;is analyzed-to demonstrate _that

-following various RCCA 'misoperation events - such as- dropped Erod(s)/ bank _or statically. misaligned rods, that'the minirum DNBR remains ~above the limit:value. Neither the primary n~or the ' secondary systems reach the reduced 1 safety set point during. this ~ event. Thus, the results of.-this analy' sis are-unaffected by increasing the tolerance on the MSSV's.to +/- 3% -

and:the-conclusions-in the RTSR remain valid.

L L2 _. _ . - - _ . - - - - _ . _ _ _ , . ~- .. -

~

ENGINEERS S erial _2 3 9 7 8 3 4 Engineer 6.6 Muaw TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLL;RANCE INCREASE Tab Page 5 of 22

4. Uncontrolled Boron Dilution (RTSL Gection 15.9,4)

This condition II aver.t is analyzed for all six modes of operation. This analysis demonstrates that sufficient negative reactivity exists, such that, should a dilution event occur, there is sufficiant time following an alarm to allow operator detection and Larmination of the event prior to a complete loss of shutdown margin and return to criticality.

The Mode 1 dilution analysis is bounded by the RCCA withdrawal at power event (RTSR 15.2.2, see item ?) while the Mode 2 dilution analysis continues to be bounded by the RCCA withdrawal at hot zero power (RTSR 15.2.1, see item 1). The MSSV set point relaxation for these events hac already been addressed. For the dilution analyses per formed in Modes 3 through 6, since adequate operator action time is assured prior to reaching criticality, ne additional heat is added to the core and no pressurization of the primary or secondary systems occurs. Changes in the MSSV set point tolerances will have no effect on the calculated available operator action time. Thus, the results of this analysis are unaffected by increasing the tolerance on the MSSV's to +/ - 3% and the conclusions in the RTSR remain valid.

5, Partial Loss of Forced Reactor Coolent Plow (RTSR Section 15.2.5)

This condition Il event is analyzed under full power conditions assuming thst 1 of 7 operating reactor coolant pumps coasts down. Tr.o reactor is promptly tripped on low reactor coolant loop flow. The analysis demonstrates that the minimum DNBR remains above the limit value. The RCS pressure increases above the-initial value during the event yet never reaches the reduced safety valve cet point. The MSSV's are not actuated during the simulation of this event. Note that no credit is taken for the observed HCS pressure rine in the DNB analysis. Thus, the results of this analysis are unaf fected by increasing the tolerance on the MSSV's to +/- 3%

and the conclusions in the RTSR remain valid.

6. Startup of an Inactive Reactor Coolant Loop (RTSR Section 15.2.6)

This condition II event is analyzed assuming a maximum initial power level consistent with 2 loop operation and the P-8' set point. The startup of an inLctive loop results in a

. _ _ ._ _ _ . _ - _ _ - - . .__.- _ _ _ _ . _ _ = - . ~ _.- - - -.

~~

. ..L <

~  ;.

l I' ENGINEERS Serial-239-02-783( l Engineer 6,4 N e o-TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page'6 of 22 reactivity insertion since the inactive loop fluid is at a lower temperature than the rest of the core. The analysis demonstrates that the minimum 'DNDR remains above the limit value.. The RCS pressure increases above the initial value yet never reaches the reduced safety valve set point. The MSSV's are not actuated during the simulation of this event. Tr.us ,

the results of this analysis are unaffected by increasing the tolerance on the MSSV's to +/- 3% and the conclusions in the RTSR remain. valid.

I

7. I4ss of External Electrical Load and/or Turbine Trip (RTSR' Section 15.2.7)

The-analysis presented in the RTSR represents a complete loss.-of, steam load from full power without a direct reactor ---

trip.e Four cases are analyzed, maximum and minimum feedback, with and without pressure control. The analysis demonstrates that, with the power-mismatch between the core'and turbine, the primary and secondary system pressures remain below 110%

of_ design _and that the minimum DNBR remains above the limit value. A sensitivity analysis was performed using the LOFTRAN computer code' _ assuming the PSRV and MSSV characteristics

' discussed in the introduction. The peak pressurizer pressure

- was- calculated , to be 2636 psia -for the minimum. feedback s

without. pressure control, case. The-peak secondary pressure; was _ calculated to be 1271 psia for all four cases. Thus:both the primary and secondary pressures continue to remain below <

--110%:of design and'the minimum-DNBR continues to remain above the limit value. -Should'the MSSV's actuate at a pressure 3_%

lower than nominal, -adequate relief capacity exists to prevent over-pressurization _of the secondary side. Thus, the_results of this analysis _are unaffected by increasing the tolerance on

'the MSSV's to_+/-13% and-the conclusions in the RTSR remain

' valid.

, 8. Loss of Normal Feedwater (RTSR Section 15.2.8)- i

'The: analysis presented in the RTSR represents a-complete:

loss of feedwater from full power. The loss of.the secondary L

side heat sink results-in a heatup-and pressurization of the ,

primary and secondary systems. The analysis-demonstrates that -

adequate emergency ~ feedwater flow' is ' delivered to the i steam n generators to remove decay heat-such that over-pressurization i

of:the primary.and-secondary-systems will not occur and;the pressurizer-doesLnot fill. Should the MSSV's-actuate at,a l'

- ,_ ~ . ~ ~ . . - -. -. - .. - _.. - . - - - - . . .~ .. -. _ - _ - ~~.

i.

. . . :l 4, ;

ENGINEERS Serial 239-02-7834 Engineer 4.6. W ,m TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 7 of,_12 lift set-point up to 3% below nominal, the maximum secondary and- primary side temperatures will be beneficially reduced.

Thus, the results of this analysis are unaffected by

-increasing the tolerance on the MSSV's to +/- 3% and the conclusions in the RTSR remain valid.

9. Loss of Offsite Power to the Station Auxiliaries (Station Blackout) (RTSR'Section 15.2.9)

The. analysis presented in the RTSR represents a complete loss of-power to the plant auxiliaries, i.e., the reactor coolant pumps, condensate pumps, etc.,-from full power. The loss of-power results in a heatup and pressurization of the primary and secondary systems. The analysis demonstrates that adequate emergency feedwater flow is delivered toithe steam generators to remove decay heat such that DNB will not occur, and the precsurizerf does not fill. Shou}d the MSSV's actuate at a lift set ._ point up to 3% below nominal, the maximum secondary and primary side temperatures will be beneficially .

reduced. Thus,- the results of this analysis are unaffected by increasing the tolerance on-the MSSV's to +/- 3% and the 1 conclusions in the RTSR remain valid.

10.- Excessive-Heat' Removal Due to Feedwater System Malfunctions-L l(RTSR Section 15.2.10)

=The analysis presented in the RTSR illustrates the plant response to a 250 % step increase in the Lfeedwater' flow to one steam generator from ' full power, ,and a: step: increase in

feedwater finw to one.. steam generator ~ at zero. power. - The
analysis demonstrates that l from zero - power- the - reactivity transient, and-thus the. minimum DNBR, is-bounded by the rod withdrawal from.sub critical event. For the full power--case, the minimum ^DNBR is~shown-to> remain,above.the' limit value.

The MSSV's _ are not actuated during this ' event even if tho _MSSV

~

lift-set point =is reduced,by up to 3%.. Thus,-the results of -

this analysis are unaffected'by increasing the tolerance on-the-MSSV's to +/- 3%'and-the conclusions in.the RTSR remain

. Valid.

11. _ Excessive Load Increase Incident (RTSR Section 15.2.11)

The - analysis presented in the RTSR describes plant

. response to a 10% step increase in load. Four different cases p--m eso,

, - . . . - ~ . . ..---. .-- .-- .. - - - . - . . . - - ~ - -

ENGINEERS Serial 239-02-7834 Engineer d. 6. C.a TECHNICAL WORK RECORD Date 1/21/91-Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 8 o f_22 are analyzedt minimum and maximum feedback, with and without reactor control. For each case it is shown that the minimum DNBR remains above the limit value. The cases'which assume no reactor control result in an RCS depressurization as the heat extraction from the-secondary side increases. The cases which take'-- credit - for reactor control result - maintain the RCS pressure at essentially the initial value. Since an increase in load results in a secondary side pressure reduction the MSSV's are not actuated. Thus, the results of this analysis are unaffected by increasing the tolerance on the MSSV's to

+/- 3% and the-conclusions in the RTSR remain valid. l l

, l 12 . : Accidental Depressurization of the Reactor Coolant System (RTSR Section 15.2.12)

For this condition II event, the-transient is initiated i by the opening of a single pressurizer relief or safety valve I

.at full power. Initially, the RCS pressure drops rapidly l until pressure reaches the hot leg saturation pressure. At  ;

this_ time the pressure decrease . continues but at a slower j

,- -rate. The analysis demonstrates that the minimum DNBR remains above the limit value. This event docs not pressurize the secondary side. -As a result, the MSSV's-are not challenged.-

Thus, the results of this analysi's are unaffected by increasing the tolerance on the - MSSV's to +/- 3% and the conclusions in the RTSR remain valid.

13. ' Accidental Depressurization of the Main Steam SystemL (RTSR ,

Sent lon = 15. 2.13 )

_ :For_this Condition II event, the. transient is initiatodi

.by the full-- opening, of a single steam dump, relief, or' safety

-valve at zero power. The anal.ysis confirmd that the minimum DNDR rema' ins above the-limit value. Since the secondary side- 3 prensures drop immediately following initiation of. the event, y the-- _MSSV's . arn ..-not actuated.- :Thus, the results of ~ _ this-analysis are unaffected by increasing . the tolerance ono the-MSSV's.to +/- 3% and the conclusions-in the RTSR remain valid.

14. Spurious Operation of the Safety Injection System at Power

_(RTSR Section 15.2.14)

For this Condition _II event, a: spurious-Safety Injection

-Signal (SIS)_ is -assumdd to be generated at full power. The m 39 qw == 1 g

1

f-ENGINEERS Serial 239-02-7834 E n g i f. e e r & A .id. u _3 TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 9 of._22 injection of borated water into the RCS reduces core power, temperature, and- pressure until the reactor trips on low pressurizer pressure. The power and temperature reduction causes a similar reduction in pressure on the secondary side.

Since the secondary side pressures drop immediately following initiation of the event, the MSSV's are not actuated. Thus,

, the results of this analysis are unaffected by increasing the tolerance on the MSSV's to +/- 3% and the conclusions in the RTSR remain valid.

15. Minor Secondary Side Pipe Breaks (RTSR Section 15.3.2)

This Condition III event continues to be bounded by the analysis presented in RTSR Section 15.4.2 -(see items 19 and 20 below).  ;

1

16. Inadvertent Loading of a Fuel Assembly into an Improper Position:(RTSR Section 15.3.3)

For the r"ent presented in the RTSR, the loading of a fuel ascombly ;ato an improper-pos.ition would affect the core power shape. Since_the power-shape and not the total power generated would be affected, the steam system conditions will remain 1 unaffected such that the MSSV's would not be affected.

Thus, the results of this analysis are unaffected by increasing _' the tolerance on the MSSV's to +/- 3% and the conclusions in the RTSR remain valid.

17. ' Complete Loss of Forced Reactor Coolant Flow (RTSR Section 15.3.4)_

This Condition III event is analyzed under full-power conditions assuming 3 of. 3. operating reactor coolant - pumps coast down. The reactor is assume $ to trip on an undervoltage signal. The analysis demonstrates' that ' the minimum . DNBR remains above tho. limit value. In the DNB analysis, no credit is taken . for the increase ~ in . prem.ure. The RCS pressure increases above-the initial ve. loa during the event yet never reaches the safety valve sot point. The MSSV's are not-actuated during this event. Thus, the results of this analysis ' are unaf fected 'by_ increasing the tolerance on the MSSV's to +/- 3% 'and the conclusions in the RTSR remain valid.

. , _ ~ - _ . - . .. . -.. - - - ~ - . . ~ . - - _ - . . -- . . . - .~ ----.

> 4

..: a ENGINEERS Serial 239-02-7834 Engineer 6.6Mm _

TECHNICAL WORK RECORD Date 1/21/91 Project Title. MSSV SET POINT TOLERANCE INCREASE Tab Pagw 10 o f_iL2 18 Single' Rod-Cluster Control Assembly (RCCA) Nithdrawal at Full Jpower (RTSR Section 1S.3.6)

For this Condition III-event,.two cases are analyzed and presented in the RTSR: automatic and manual reactor control. l In-both cases an increase in core power, coolant temperature, 1 and-hot channel factor result in a reduction in the minimum-DNBR. Tie._ analysis demonstrates that, although it is not possible for all cases to ensure that DNB will not occur, an ,

' upper _ bound on the number of fuel rods experiencing DNB-is less than or equal to 5%. Since this event is a limiting _DNB event and not peak pressure limiting, credit is not taken for any pressure increase associated with this event. The MSSV's are not actuated during this event. - Thus, the results of this ,

analysis are ~ unaffected by increasing the tolerance on the l

= MSSV's to +/- 3% and the conclusions in the RTSR remain' valid.- l

~19. Rupture of a Main Steam Line (RTSR Section 15.4.2.1) '

For this' Condition IV event, the transient is assumed to be initiated by the instantaneous double-ended rupture of a main steam 'line. Since the , secondary side pressures drop immediately- following initiation of the event, the MSSV's are not actuated. Thus, _the results of' this analysis are

_ unaffected by increasing the tolerance on the MSSV's to +/- 3%

and the conclusions in the RTSR remun valid.
20. Rupture of a' Main Feedwater pipe _(RTSR Section.15.4.2.2)

For this Condition IV event, the double-ended rupture of

.t Inain feedwater pipe' initially results in a cool'down of,the -

RCS due to the heat removal of t% steam generator blowdown.

This _ cool down-period is followed'by-a heat ~up.as the high levels ' of - decay -heat. and the lack of inventory . on 'the secondary side.results in' inadequate heat transfer.: The event-is analyzed to show that adequate'. - heat removal capability-exists-to-remove' core' decay heat and stored energy:following

.a reactor trip from full power and_that the core remains in a' coolable ; geometry. This is . accomplished by applying the

-strict criterion - that no hot 1eg boiling _ occurs during' the transient. ~For this event, the MSSV'_s are actuated during the-heatup phase following reactor' trip. A sensitivity analysis-has been performed using the LOFTRAN code . assuming .the increased.MSSV_ set points. Maximum steam system pressures were. calculated:to be 1272 psia. Minimum subcooling margin in

y 4 ENGINEERS B eri al.2 3 9-0 2-7 8 3 4 Engineer 6.6.W,u w TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT _TOLERAliCE INCREASE Tab Page_J_1_ o fl2 the RCS was found to bo 23.5 F. Thus the analysis shows that the secondary system is not over pressurized and no hot leg boiling occura in the RCS hot leg. A reduction in the MSSV set point will serve to reduce maximum secondary side temperatures and pressures. Thus, thu restilts of this analysis are unaffected by increasing the tolerance on the  !

MSSV's to +/- 3% and the conclusions in the RTSR remain valid.

21. Single Reactor Coolant Pump Locked Rotor (RTSR Section 15.4.4) l This Condition IV event is analyzed under full power i conditions assuming the instantaneous seizure of one Reactor Coolant Pump motor. This results in a rapid RCS flow ,

reduction and pressure rise with possible DND. The reactor is promptly tripped on a low flow signal. The analysis demonstrates that no more than 15% of the rods experience DNB and that the RCS peak pressure remains below that which would cause stresses to exceed the faulted condition stress limits.

The secondary system does not reach the MSSV set point during the simulation of this event. Thus, the results of this analysis are unaffected by increasing the tolorance on the MSSV's to +/- 3% and the conclusions in the RTSR remain valid.

22. Rupture of a Control Rod Drive Mechanism Housing (RTSR Section 15.4.6)

For this Condition IV event, a rapid reactivity insertion and increase in core power leads to high local fuel and clad temperatures and possible fuel and/or clad damage. Four cases are analyzed: beginning of life, end of life, hot zero power, and hot full power. The analysis shows that the fuel and clad limits discussed in RTSR Section 15.4.6 are not exceeded and that RCS pressure does not exceed the faulted condition stress limits. The MSSV's are not modeled as part of this over pressure analysis and are therefore not required to operate.

Thus, the results of this. analysis are unaffected by increasing the tolerance on the MSSV's to +/- 3% and the conclusions in the RTSR remain valid.

23. Steamline Break Mass / Energy Release -

Inside/Outside Containment Various steam line break ( iscs are analyzed for the purposes of gener atf g mass and energy release rates which are

6- ,

ENGINEERS Serial _2 3 9-02-7 6 3 4 Engin e er 6. 6. Wes TECHNICAL WORK RECORD Date_ 1/21/91 Proj ect Title _ MSSV SET POINT TOLERANCE INCREASE _ Tab Page_12_ of_22 then applied to containment response or compartment environmental analyses.

break sizes and initial Cases occurring powerare performed assuming various levels. For small breaks at high power levels, it is possible that pressurization of the primary and secondary systems may occur.

Specifically, if the energy release through the break is less than the decay heat deposition into tbo RCS, pressurization may occursince However, possibly the torelief the point of safety valve actuation.

capacity of the MSSV's is undiminished, there is sufficient prest.urization of the secondary systems. capacity to the prevent over Raising MSSV set points will have previously no impact upon the mass and energy releases calculated.

Reductions in the MSSV set points will serve to reduce and energy release rates. the primary and secondary side temperatures mass and energy releases Thus, the results of the calculated increasing the tolerance on the MSSV's to +/- 3% and the are not adversely affected by conclusions in the FSAR remain valid.

SET POINT IMPACT In addition to the impact upon the non-LOCA accident analyses, increasing the MSSV safety valve tolerance to +/- 3% will also impact protection.the core limits and the over-power and over-temperature reactor is protected As seen in Figure by the MSSV 15.1-1 line. of the VCSNS RTSR, the the steam generator, primary to The temperature drop across proportional to power. secondary, is approximately The secondary temperature is approximately constant set point.at the saturation temperature corresponding to the MSSV MSSV set point saturation temperatureTherefore the primary temperature cannot r across the steam generator) . (plus the temperature drop of the boundaries on power This temperature limit serves as one and temperature in addition to the bounds imposed bv the over~ power and over-temperat.ure trip set points.

.By increasing the MSSV set point by 3%, the saturation temperature is increased by approximately 4'F. By decreasing the MSSV-set point by 3%,

approximately 4 F. the saturation temperature is decreased by Examination of Figure 15.1-1 reveals that movement of the steam generator safety valve line by 4 F will not result in violation of core 1imits. A reduction in the saturation delta-T protection system must provide DNBThus, protection.

the temperat i

over-power provido and over-temperature protection from core l'imits.delta-T set points continue to t

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ENGINEERS Serial 239-02-7034 Engineer 6. 64.o4 4 TECHNICAL WORK RECORu Date 1/21/91 Project Title _HSSV SET POINT TOLERANCE INCREASE tab Page 13 o f __212 LQ.CA The following presents the effect of the proposed MSSV set point tolerance increase on the LOCA related analyses.

1. Large Break LOCA (FSAR Chapter 15.4.1)

The licenaing basis large break LOCA analysis for VCS9s was performed using the 1981 evaluation model with BASH. The analycis assumed a total core peaking factor (F q ) of 2.45 with uniform 15% steam generator tube plugging. The analysis determined the limiting break r.ize to be a double-ended guillotine with a discharge coefficient equal to 0.4. The peak clad temperatur e for this case was 2141 F. However due to the fuel load for Cycle 6 being a transition core with Vantage 5 fuel and Standard fuel existing in the core, a transition core penalty of 50 F was assessed. Thus the ef fective peak clad temperature was determined to be 2191'F.

The large break LOCA analysis does not model the MSSV's.

This is because the RCS is quickly depressurized below that of the steam generator secondary pressure and the MSSV's are never challenged. Thus the large break analysis results are not dependent on the performance of the MSSV's. Therefore, the large break LOCA analysis results are not adversely affected by the revised .3SV set point tolerances.

2. Small Break LOCA (FSAR Chapter 15.3.1)

The licensing basis small break LOCA analysis for VCSNS was performed using the NOTRUMP computer code. The analys.is assumed a total core peaking factor of (F of 2.50 with 15%

steam generator tube plugging. The smal'. b)reak LOCA analysis assumed the plant was operating in Mode 1 at 102% reactor power. The analysis considered break sizes of 2, 3, and 4 inch diameters and determined the limiting break size to be a 3 inch diameter break located in the cold leg. The limiting peak clad temperature was 2095 F.

The smell break LOCA analysis requires the MSSV's to remove decay heat from the RCS, Since the small break LOCA assumes loss of of site power with reactor trip, no credit is taken for operation of the steam generator power operated relief valves or the steam dump system. After reactor trip, the secondary pressure quickly reaches 1225 psia. However, after this initial spike, secondary pressure remains at the first safety lif t pressure for the remainder of the transient.

Since the tolerance for the first MSSV set point pressure

.o .

I ENGINEERS Berial 239-02-7834 Engineerd.6. h m TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 14 of 22 remains unchanged and the maximum pressure obtained is well below the maximum MSSV set point, the small break LOCA analysis is unaffected by the increase in MSSV tolerance to

+/- 3% for the four highest set pressures. Therefore, the conclusions in the FSAR remain valid.

3. Hot Leg Switchover to Prevent Potential Boron Precipitation (FSAR Chapter 6.3.2.5)

Post-LOCA hot leg recirculation switchover time is determined for inclusion in emergency procedures to ensure no boron precipitation in the reactor vessel following boiling in the core. This time is dependent on power level, boron concentrations, and water volumes of the RCS, RWST, and accumulators. Since the secondary safety valves affect neither the maximum boron concentrations nor the volumes assumed for the RCS, RWST, and accumulators, there is no effect on the post-LOCA hot leg swit,: hover time.

4. Blowdown Reactor Vessel and Loop Forces (FSAR Chapter 3.9.3)

The blowdown hydraulic loads resulting from a loss of coolant accident are considered in section 3.9.3 of the VCSNS FSAR. Because the maximum loads are generated so quickly, a change in the secondary safety valve set point tolerances would have no ef fect on the analysis results. Thus, it can be concluded that the consequences of the blowdown reactor vessel and loop forces calculations will not be affected by the revised MSSV set point tolerances.

5. Post-LOCA Long Term Core Cooling; Westinghouse Licensing l Position (FSAR Chapter 15.4.1)

The Westinghouse licensing position for satisfying the requirements of 10CFR Part 50 section 50.46 paragraph (b) item (5) "Long Term Cooling" is defined in WCAP-8339. The Westinghouse Evaluation Model commitment is that the reactor I will remain shut down indefinitely by borated ECCS water

[ residing in the sump following the postulated LOCA and when SI l

switchover is accomplished. Since credit for the control rods is not taken for large break LOCA, the borated ECCS water provided by the accumulators and the RWST must have a boron

! concentration that, when mixed with other water sources, will

o

  • ENGINEERS Serial 239-02-7B34 l Engineere 4 & M l TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 15 of 22 result in the reactor core remaining subcritical at sudng all control rods out.

Sump boron concentration is determined by the accumulation of all potential water sources in the containment, based on each respective source boron concentration. The revised secondary safety valve set point tolerance will not affect the post-LOCA sump boron concentration. It is therefore concluded that there would be no change to the long term cooling capability of the ECCS system as a result of the revised MSSV set point tolerance.

S_ TEAM GENERATOR TUBE RUPTURE The FSAR analysis for a steam generator tube rupture (SGTR) is performed to evaluate the radiological consequences due to the SGTR event. The major factors that affect the radiological consequences for a SGTR are the amount of radioactivity in the reactor coolant, the amount of reactor coolant transferred to the secondary side of the af fected steam genera 7or through the ruptured tube, and the amount of steam released from the steam generator to the atmosphere.

A SGTR results in a decrease in pressurizar pressure due to the loss of reactor coolant inventory. Reactor trip and SI actuation were assumed to occur as a result of low pressure for the l

VCSNS SGTR analysis. A loss of offsite power was also assumed to occur at the time of reactor trip and thus, the steam dump system was assumed to not be available. The energy transfer from the primary system following reactor and turbine trip causes the secondary side pressure to increase rapidly after reactor trip until the steam generator power operated relief valves (PORV's) and/or safety valves lift to dissipate energy. For the SGTR l alialysis in the VCSNS FSAR, it is assumed that the secondary pressure is maintained at the lowest secondary safety valve (MSSV) set point following reactor trip. After reactor trip and SI initiation, the RCS pressure was assumed to reach equilibrium at the point where the incoming SI flow rate equals the outgoing break flow rate, and the equilibrium pressure and break flow rate were assumed to persist until 30 minutes after the accident.

Since the equilibrium break flow rate is a function of the primary to scnndary pressure differential, a change in the MSSV l

set point Palerance to +3% will result in the secondary pressure being maincained at a :igher pressure during this 30 minute period

i + .

O ENGINEERB Serial 239-02-7834 Engineer 61 W .m TECHNICAL WORK RECORD Date 1/21/91 Project Title._tiSSV SET POINT TOLERANCE INCREASE Tab Page 16 of 22 thereby decreasing the primary to secondary pressure differential.

This will result in an decrease to the primary to secondary break flow and thus, a slight decrease in the atmospheric steam release via the ruptured steam generator. Therefore, for a positive increase in set point tolerance (+3%), the FSAR analysis results would remain conservative.

A change in the MSSV set point tolerance to -3% will result in the secondary pressure being maintained at a lower pressure during the 30 minute period thereby increasing the primary to secondary pressure differential. This will result in an increase to the primary to secondary break flow and the atmospheric steam release via the ruptured steam generator.

It is noted that several safety evaluations for plant changes at VCSNS have been previously performed by Westinghouse. The plant changes include changes to secondary operating level, changes to pressurizer operating level, fuel changes, temperature variations associated with the margin broker program,15% steam generator tube plugging, and increased high head safety injection flow.

Sensitivity studies were performed to determine the bounding primary to secondary break flow and atmospheric steam release via the ruptured steam generator for these changes and the increased MSSV set point tolerance. The results of these sensitivity analyses indicate that the primary to secondary break flow would increase but remains less than the reported result due to conservatism in the VCSNS FSAR analysis. The atmospheric steam release via the ruptured steam generator would be increased by approximately 32% over the result reported in the VCSNS FSAR.

Finally, it is noted that the reactor coclant activity assumed for the SGTR analysis in the VCSNS FSAR is based on 1% fuel defects and is assumed to be independent of the transient conditions; therefore this assumption would not be affected by the aforementioned changes.

An evaluation incorporating these bounding mass release results was completed to determine the impact on the offsite radiological doses reported in the VCSNS FSAR for the SGTR event.

The results of the analysis indicate that the whole body dose reported in the FSAR remains bounding. However, the thyroid dose will increase by 24% over the approximate 0.43 rem reported in the FSAR. Although these results show an increase in the thyroid dose over those presented in the FSAR, this does not constitute an increase in the consequences of the accident. This judgement is based on the NSAC 125 " Guidelines for 10 CFR 50.59 Safety Evaluations" criteria for " increases in consequences," 1.e.; the dose increases are small and the total dose is very low, being well

i e .

o

  • ENGINEERB Serial 239-02-7834 Engineer M. W w _

TECHNICAL WORK RECORD Date 1/21/91 Proj ect Title MSSV SET POINT TOLERANCE INCREASE - Tab Page 17 of_22 within the NRC definition of a "small fraction" of the 10CFR100 exposure guidelines. This "small fraction" is defined as 30 m thyroid and 2.5 rem whole body which is 10% of the 10 0FF 'O guideline values of 300 rem thyroid and 250 rem whole body.

GILBERT SAFETY AJALYSES While the Westinghouse analyses had no problem with the increased set point tolerance for all 15 MSSV's, the Gilbert analyses would allow the increased tolerance on all but the lowest set bank of MSSV's. The lowest set bank of MSSV's are relied upon to be at set point +/- 1% for two reasons. The first is an operational rather than a safety concern. If the lowest set bank of valves were to be set with a - 3 % tolerance, the blow down closure point would be lower than the no load Steam Generator pressure of 1092 psig. The second reason is to maintain a +/- 1%

tolerance on set point so as not to affect the capability of the Emergency Feedwater (EFW) system to provide the required water to the Steam Generators.

The EFW flow calculations are based on the set point of the lowest set bank of MSSV's. Westinghouse states that for the worst case transient, the design flow rate of 82.3% of full rated flow 0 100% reactor power will occur for a short period of time. This is because a reactor trip occurs very quickly in the transient. Once a reactor trip occurs, no additional heat (except for decay heat) is added to the system and secondary system pressure and flow rate quickly decrease from that point on. This is exemplified by the fact that secondary steam flow through the MSSV's never goes above 82.3% of full flow even though the reactor was at 100% power upon initiation of the event. Once the pressure transient has turned around, pressure will quickly drop allowing the MSSV's to close and the requirements for EFW come into play. At this time, since little or no heat is being input to the secondary system, pressure Will not get above the set pressure of the first bank of MSSV's.

Therefore, the EFW flow calculations are based on the set point of the lowest set bank of MSSV's. Note: With the increased set point tolerance, it is possible for the second bank of MSSV's to be set slightly below (10 psi) the first bank of MSSV's. This was analyzed by Gilbert and determined to have an insignificant effect on the EFW flow split, i.o; the flow split still meets the design criteria. Since EFW flow capability still meets the design

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ENGINEERS Berial 239-02-783L Engineer A 4. Mu %

TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 18 of 22 criteria, there is no effect on the capability of the EFW system to l mitigate accidents.

  • CONCLUSIONS The capability of the MSSV's to mitigate secondary system over-pressure events is maintained with the increased set point tolerance. Therefore the consequences of an event affected by the MSSV's are not increased. The probability of a MSSV malfunction is not affected by this change. No new previously unanalyzed accidents or equipment malfunctions are introduced by this change.

The Tech. Spec. margin of safety remains unchanged for the increase of MSSV set point tolerance from +/- 1% to +/- 3%. The Tech. Spec. requirements for MSSV set point are based upon the requirements of the ASME B&PV Code Section III. Westinghouse safety analyses, in accordance with the ASME B&PV Code, have shown that anticipated transients coincident with an increased MSSV set point tolerance (+3%) will not cause the secondary system pressure allowables (110% design pressure) to be exceeded. The Tech. Spec.

margin of safety is that above the ASME limit of 110% design pressure and below ultimate failure. Thus the Tech. Spec. margin of safety for maximum secondary pressuro remains unchanged.

The EFW flow calculati)ns are based on the set point of the lowest set bank of MSSV's. Therefore, the set point tolerance for the lowest set bank of MSSV's will not be changed. Since EFW flow capability will remain in compliance with the design criteria, Tech. Spec. margin for EFW flow will remain unchanged.

l l

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o 4 ENGINEERS Serial 239-02-7334 Engineer 6,6. V,u..w -

TECHNICAL WORK RECORD Date 1/21/91 _ . _

Pro 4,ect Title MSSV SET POINT TOLERANCE INCREASE Tab Page_J,jl, o f 2 2 The new Main' Steam Safety Valve set point tolerances are as follows:

MAIN STEAM SAFETY VALVEB PROPOSED-MAIN STEAM MAIN-STEAM MAIN STEAM MBBV SET BET POINT LOOP A LOOP B LOOP C POINT TOLERANCE XVS-2806 A XVS-2806 F XVS-2806 K 1176 PSIG +/- 1%

XVS-2806 B XVS-2806 G XVS-2806 L 1190 PSIG +/- 3%

XVS-2806 C XVS-2806 H XVS-2806 M 1205 PSIG +/- 3%

XVS-2806 D XVS-2806 I XVS-2806 N 1220 PSIG +/- 3%

XVS-2806 E XVS-2806 J XVS-2806 P 1235 PSIG +/- 3%

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a 4 l EN3INEERS Scrici 239-02-7834 >

. Engineer 4,4 C 4%  !

TECHNICAL WORK RECORD Date 1/21/91 ]

Proj ect Title, MSSV SET POINT TOLERANCE INCREASE Tab Page 20 o f_21 i 4

10CFR50.59 SAFETY EVALUATION QUESTIONS J

XCli ILQ 1

1. May the proposed activity increase the probability of __

X occurrence of an accident previously evaluated in the FSAR or FPER?  !

Basis: The MSSV's orovide orotection from over-oressurizption of the secondary systems and are actuated af ter an accident is initiated. The accidental depressurization of the Main Steam system events can be initiated by the openina of a MSSV. Increasina the tolerance on these valves does not create a new failure mode or result in a lift set ooint that would increase the probability of an inadvertent onenina of

+hese valves.

XS.fi 11 9

2. May the proposed activity increase the consequences of ___

X an accident previously evaluated in the FSAR or FPER?

Basis: The capability to mittaate over oressure events =femains within accentable limits (110% of desian Dressure maximum) . (

MSSV rated flow remains unchanced. The canability of the EFW system to miticate events remains unchanced because the EFW system continues to meet the desian criteria. As pf_eviously discussed. DNBR and PCT values affected by the Non-LOCA accident events remain within the limits specified in the licensina basis documentation. It has been demonstrated that the mass /eneray releases inside and outside the containment oreviously documented in the FSAR remain valid. In addition, a review of the SGTR analyses chows that an increase in safety valve set ooint tolerance will decrease the primary to seconds.rv oressure dif ferential I and decrease the break flow rate. The steam release from the ructured steam cenerator would decrease slichtly, and will have an insianificant chance in the offsite doses.

However, a decrease in safety valve set point tolerance will increase the crimary to secondarv oressure differential and increase the break flow rate. As a result, the nte02 release from the ruptured steam cenerator would increase Verification Approval Type of Verification Verifier Signature /Date Signature /Date ohny L cT j.;, y

, s. - ,,

t  %

ENGINEERS ~ Scrici 239-02-7934, y Engineer 4.4 M wa .s TECHNICAL WORK RECORD Date 1/21/91 Project Title MSSV SET POINT TOLERANCE INCREASE Tab P a g e _ 2.1,,,, o f 2 slicht), causina a small increase in the offsite thyroid

' doses. This-increase in dose under NSAC 125 cuidelines is not considered to constitute an increase in the consecuences of the accident because the revised doses are still well within the current NRC acceptance criteria as set forth in the Standard Review Plan.

XRa M

3. May the proposed activity increase the probability of _2L occurrence of a malfunction of equipment important to safety previously evaluated in the FSAR or FPER?

Basis: The orobability of a failure of an MSSV is not affected by the increase in set noint tolerance to +/- 3%. By maintainina the system upset- desian oressure. the orobability of a malfunction of any other eauioment imoortant to safety is not increased.

Yaa M

4. May the proposed activity increase the consequences of X a ' malfunction of equipment important to safety previously evaluated in the FSAR or FPER?

Basis: As oreviously discussed, the canability to mitiaate over Dressure events remains within accentable limits (110% of desian oressure maximum). MSSV rated flow remains unchanced. The capability of the EFW system to mitiaate events remains unchanced because the- desian criteria continues to be met. DNBR and PCT values affected by the Non-LOCA and LOCA accident events remain within the limits soecified in the licensina basis documentation. It has been demonstrated -that the mass /enerav releases -inside and outside the containment oreviously documented in the FSAR remain valid. Althouah it is in reality a failure of eauipment, the SGTR event is considered to be an accident, ag_such, the ef fects of a chance in MSSV set ooint tolerance on-the SGTR analyses was discussed--in auestion 2. The chance in MSSV set ooint does not imoact the ability of any other safety nystem to perform its intended safety function.

Verification Approval Type of Verification Verifier Signature /Date Signature /Date 4 rw WW M'{l

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+ 4 80rici 239-02-7834 ENGINEERS

,. Engineer 6.4 Mste w ,

TECHNICAL WORK RECORD Date 1/21/91 I Project Title MSSV SET POINT TOLERANCE INCREASE Tab Page 2 2 . o f._22 c lene -@

. 5. May the proposed activity create the possibility of an X accident of a 'different type than - any previously ovaluated in the FSAR or FPER? l 1

Basis: As oreviousiv stated, the' MSSV's orovide over-oressurization orotection for the secondary system. The analyses results as oresented in the FSAR remain valid and no new failure mechanisms were determined. Thus, the possibility of an accident which is different than any already evaluated in the FSAR and would not be created as a result of increasina the tolerance on the four hiahest MSSV set oressures to +/-

1.L.

6. May the proposed activity create the possibility of a X different type of malfunction of equipment important to safety than any previously evaluated in the FSAR or FPER?

Basis: Allowina a laraer MSSV set ooint tolerance for the four hiahest set pressures does not result in any conditions beina chanaod which could result in the malfunction of eauioment imoortant-to safety different from any evaluated in the PSAR/FPER. ,

Yes M

7. Does the proposed activity reduce the margin of. safety X as- defined in the basis. for any technical specification?

. Basis: As indicated in the above evaluation, the conclusions Drovided in the FSAR remain valid. All acceptance criteria

.Q_ontinue to= be met. Therefore, there is no reduction in the marain ' of safety defined in the bases for any Technical

, Specification.

1 Verification Approval Type of Verification Verifier Signature /Date Signature /Date hd yYM:lc"W k-fd C

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