LD-84-070, Forwards Proposed Changes to CESSAR Re Startup Testing for Review.Changes Currently in FSAR Change Process & Will Be Incorporated in CESSAR Amend 10

Forwards Proposed Changes to CESSAR Re Startup Testing for Review.Changes Currently in FSAR Change Process & Will Be Incorporated in CESSAR Amend 10

ML20100K444
Person / Time
Site:	05000470
Issue date:	12/05/1984
From:	Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
LD-84-070, LD-84-70, NUDOCS 8412110116
Download: ML20100K444 (35)

Text

C-E Power Systems Tel. 203/688-1911 Combustion Engineering, Inc. Telex: 99297 1000 Prospect Hill Road Windsor, Connecticut 06195 POWER M SYSTEMS STN 50-470F December 5, 1984 LD-84-070 Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

CESSAR Amendment 10

Dear Mr. Eisenhut:

As a result of startup testing conducted at Palo Verde on the first System 80" plant, C-E has identified changes to CESSAR. These proposed changes are described in the attachment and are provided for NRC Staff review.

These changes are currently in the FSAR change process and will be added to CESSAR in Amene',ent 10. If you have any questions or comments concerning these changes, please feel free to call me or Mr. T. J. Collier of my staff at (203) 285-5215.

Very truly yours, COMBUSTION ENGINEERING, INC.

ex L E. Wherer Director Nuclear Licensing AES:las Attach.

cc: P. Moriette 3

8412110116 841205 PDR ADOCK 05000 hg

^ r,

LD-84-070

- Attachment h Page-1 of 2 L

SUMMARY

OF CHANGES Chapter 5: Reactor Coolant System and Connected Systems Section 5.2.2.10 and F1'gure 5.2-2 are changed. to reflect a lower maximum &T across a steam generator. This change -is made to allow additional operational flexibility and limits the maximum 6T at which a Reactor Coolant Pump can be L started to 100 F. Additionally, clarification is added to more accurately reflect the Shutdown Cooling System relief valves used in construction.

Figure 5.2-1 is modified to credit the typical difference in height between the-pressurizer and the shutdown cooling system. This change was also made to provide operational flexibility.

Section 5.4.2.4.1 is changed to correct the Section numbers referenced. This change is considered editorial in nature.

Section 5.4.10.3 is revised to indicate that pressurization rate testing is not performed-during Hot Functional Testing. Pressure control setpoints are 1 determined analytically and checked during Power Ascension Testing.

Chapter 6: Engineered Safety Features Tables 6.3.3.3-1 and 6.3.3.3-2, and Figure ' 6.3.3.2-5L, are modi fied to ease Low Pressure Safety Injection flow requirements. These flow changes were found to

'be sufficient to preserve the present ECCS performance results. Compar*1oa of

-the revised figure with the figure currently ir. CESSAR demonstrates the insignificance of this change on the worst case postulated break.

Appendix 6B: Iodine Removal System Section 7.16.4 is revised to remove unwarranted restrictions on the transfer of hydrazine and to clarify the qualifications of the arrangement used to transfer hydrazine to ke Spray Chemical Additional Tank.

i Chapter 7: Instrumentation and Controls Section 7.1.2.10 is revised to clarify conformance to IEEE 384-1974 as augmented by Regulatory Guide 1.75 (Rev. O, 2/74). It shows that the ccarnitment to perform specific analyses have been completed.

LD-84-070 4

Attachment .

Page 2 of- 2. l l

a. l Chapter 9: ' Auxiliary Systems

' Tables 9.2-1 and ' 9.3-1,' and Section - 9.3.4.1.3.2, are updated to reflect C-E's latest guidance on chemistry controls.

Chapter 10: Steam and Power Conversion System ,

~ Tables 10.3.4-1 and 10.3.4-2, and Section 10.3.4, are-updated = to r' e flect .C-E's latest guidance o,n chemistry controls. . ,

1 .

s . .

Chapter 14: LInitial Test Program' Section 14.2.12.2.5 (page 14.2-59) is revised to correct an editorial error.

The letdown system valves tested are the control valves not the-isolation valves.

Chapter 16: Technical Specifications

' Specifications 3.4.1.3 and 3.4.1.4.1 and their bases are modified to reflect

! the reduced steam generator 6T described under Chapter 5 above.

Appendix A

.The Regulatory Guide 1.68.2, Revision 1 position statement is revised to reflect compliance with the intent of that guide. This change was effected in Amendment 6 to.CESSAR dated November 20, 1981, but was not reflected in Appendix A.

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5.2.2.10.1.1 Credit for Operator Action No credit is taken for operator action for 10 minutes after the operator is I made aware that a transient is in progress.

5.2.2.10.1.2 Single Failure In the LTOP mode, each SCS relief valve is designed to protect the reactor vessel given a single failure in addition to a failure that initiated the prwssure transient. The event initiating the pressure transient is considered to result from either an operator error or equipment malfunction. The SCS relief valve system is independent of a loss of offsite power. Each SCS relief valve is a self actuating spring-loaded liquid relief valve which does not require 4.

control circuitry. The valve opens when the RCS pressure exceeds its setpoint.

The redundant SCS suction line trains between the RCS and SCS relief valves meet the single failure criteria as described in paragraph 5.4.7.1.2 and table 5.4.7-3. No single failure of an isolation valve or its associated interlock will prevent one relief valve from performing its intended function.

5.2.2.10.1.3 Testability Periodic testing of the SCS isolation valves is defined in the Technical Speci -

l fications, paragraph 16.3/4.5.2.

5.2.2.10.1.4 Seismic Desion and IEEE 270 Criteria i The SCS suction line relief valves, isolation valves, asso-f ated interlocks, and

( instrumentation are designed to Seismic Category I requiren.2nts as discussed in

- subsections 3.2.1, paragraph 5.4.7.2.5 and table 3.2-1. The interlocks and instrumentation associated with the SCS suction isolation valves satisfy the appropriate portions of IEEE 279 criteria as discussed in paragraphs 5.4.7.2.5, 7.6.2.1.1 and 7.6.2.2.1.

5.2.2.10.2 Design and Analysis

{ In demonstrating that the SCS relief valves meet the criteria listed in paragraph 5.2.2.10.1', the following additional information is provided.

l 5.2.2.10.2.1 Limiting Transients Transients during the low temperature operating mode are more severe when the RCS is operated in the water-solid condition. Addition of mass or energy to an isolated water-solid system produces increased}f" system pressure. The severity l of the pressure transients depends upon the rate ~and total quantity of mass or energy addition. The choice of the limiting LTOP transients was based on evalu-ations of potential transients for System 80 plants. The most limiting transients initiated by a single operator error or equipment failure are:

a) An inadvertent safety injection actuation (mass input).

b) A reactor coolant pump start when a positive steam generator to reactor vessel a T exists (energy input).

5.2-6a

The transients were determined as most limiting by conservative analyses which maximize mass and energy additions to the RCS. In addition, the RCS is assumed to be in a water-solid condition at the time of the transient; such a condition has been noticed to exist infrequently during plant operation since the operator is instructed to avoid water-solid conditions whenever possible.

Figure 5.2-1 shows the result of the inadvertent safety injection actuation transient analysis when the RCS is in the LTOP mode. The mass addition due to the simultaneous operation of two HPSI and three charging pumps was considered, along with the simultaneous addition of energy from decay heat and the pressurizer heaters.

., _ ke hggy se ac.; too,F Figures (5. 2-2+ ehto.1shows the rbsult of the transient analysis of a re tor coolant pump stjrt when a steam generator to reactor vessel a T of

^

exists. This a T is _r-Wc ed ;rcate- t": : y t':t a ; =:S' ==r " the :ytt-- during the LTOP mode. In addition to considering the energy addition to the RCS from the steam generator secondary side, energy addition from decay heat, the reactor coolant pump and all pressurizer heaters were also included. In this analysis the steam geneators were assumed to be filled to the zero power, normal water level. For conservatism, the secondary water, both around and above the U-tubes, was assumed to be thermally mixed in order to maximize the energy input to the primary side. This assumption is conservative since as a result of the temperature distribution within the steam generator during the transient, the water inventory above the tubes is practically isolated thermally from the heat transfer region.

Therefore the heat transfer rate, and thus the primary side pressure, is not sensi-tive to the secondary side water level as long as the tubes are covered.

tco*F ioo*P On the basis of exper ence, the AT value of -MF+ used in the analysis is much larger than any AT that might be expected during plant operation. This maximum allowable AT of -MG2# will prevent pressurizer pressure from exceeding the minimum P-T l imi t ' '---^ ' " ' " - allowed for the lowest system temperature during the LTOP mode of operation. (See Technical Specification Figure 3.4-2). During RCS cooldown using the shutdown cooling system, coolant circulating with the reactor coolant pumps serves to cool the steam generator to keep the temperature differ %pe between the reactor vessel and the steam generator minimal. Procedures l will direct the operator to maintain the aT below approximately 20 F.

LTOP transients have not been analyzed for the simultaneous startup of more than one reactor coolant pump (RCP). Such operation is procedurally precluded since the operator starts only one RCP at a time and a second RCP is not started until system pressure is stabilized. Additionally, there is an LTOP transient alarm that should indicate that a pressure transient is occurring. Accordingly, the second RCP would not be started. >

He.3/4.4.1.3 Technical g Specification section IS.2/ * '.: requires that the operatornot start an RCP ir the aT exceeds-!SO'". However, as mentioned above, administrative procedures will ensure t at the AT is maintained below approximately 20*F.

100*F The results of the analyses provided in Figures 5.2-1 and 5.2-2 show that the use of either SCS relief valve will provide sufficient pressure relief capacity to mitigate the most limiting LTOP events identified above.

5.2-6b

5.2.2.10.2.2 Provision for Overpressure Protection During heatup, RCS pressure is maintained below the maximue pressure for SCS operation until RCS cold leg temperature exceegs the applicable P-T operating curve temperature corresponding to 2500 lb/in. a (see Figure 3.4-2 in the Technical Specifications). If SI-651 and 653 or SI-652 and 634 SCS suction isolation valves are open and RCS pressure exceeds the maximum pressure for SCS operation, an alarm will notify the operator that a pressurization transient is occurring during low temperature conditions. Either SCS relief valve will terminate inadvertent pressure transients occurring during RCS temperature below ghe applicable P-T operating curve temperature corresponding to 2500 lb/in. a. Above the maximum LTOP temperature, overpressure protection is provided by the pressurizer safety valves when the SCS relief valve is isolated from the RCS.

During cooldown whenever RCS cold leg temperature is below the applicable temperature for LTOP, the SCS relief valves provide the necessary protection.

If the SCS is not aligned to the RCS before cold leg temperature is decreased ,

to the maximum to open te:aperature the SCS suction requiring isolation valves LTOP, (SI-651,an alarm will 652,153, 654notify)the operator

. The maximum temperature requiring LTOP is based upon the evaluatian )f the applicable P-T curves. However, the SCS can not be aligned to the RCS antil the pressure is -

below the maximum pressure allowing SCS operation (see, paragraph 5.4.7.2.3, itema.2).

These LTOP conditions are within the SCS operating range. Technical Specifica-tion section 16.3/4.4.$.3 requires the SCS suction line isolation valves to be l t open when operating in the LTOP mode. Also, this Technical Spacificatiot N ensures that appropriate action is taken if one or more S2 relief valves are out of service during the LTOP mode of operation.

l Either SCS relief valve will provide sufficient relief capacity to prevent any pressure transient from exceeding the isolation interlock setpoint (See figures 5.2-1 and 5.2-2).

5.2.2.10.2.3 Equipment Parameters The SCS relief valves are spring-loaded G b 4 liquid relief valves with l sufficient capacity to mitigate the most limiting overpressurization event.

Pertinent valve parameters are as follows:

Parameter

gerninal Setpoint 450 lb/in.2 ,gj,, f l

Accumulation 10%

Capacity 4000 (0 10% acc) gal / min Since each SCS relief valve is a self actuating spring-loaded liquid relief

. valve, control circuitry is not required. The valve will open when RCS pressure exceeds its setpoint.

5.2-6c

The SCS relief valves are sized, based on an inadvertent safety injection actuation signal (SIAS) with full pressurizer heaters operating from a water-solid condition. The SIAS assumes simultaneous operation of two HPSI pumps and three charging pumps with letdown isolated. The resulting flow capacity requirement for water is 4000 gpm. The analysis in Section 5.2.2.10.2.1 assumed that either SCS relief valve relieved water at this rate. The design relief capacity of each of two SCS relief valves (shown in P&ID Figure 6.3.2-18) as supplied by the valge manufacturer i: 5100 ;;r; "i: der'; reli:f ::;::it3 n: d htne minimum *

• required relief capacity of 4000 gpm 't' ~ufficient margin in relieving capacity for even the worst transient. The SCS refef valves are Safety Class 2, designed to Section III of the ASME Code. www( , ,J .,; u 5.2.2.10.2.4 Administrative Controls Administrative controls necessary to implement the LTOP provisions are limited to those controls that open the SCS isolation valves. Before entaring the low temperature region for which overpressure protection is necessary, RCS pressure is decreased to below the maximum pressure required for SCS operation. Once the SCS is aligned, no fJrther specific administrative procedural Controls are needed to ensure proper overpressure protection. The SCS will remain aligned whenever the RCS is at low temperatures and the reactor vessel head is secured. As designated in Table 7.5-2, indication of SCS isolation valve position is provided.

5.2.3 REACTOR COOLANT PRESSURE B0UNDARY MATERIALS 5.2.3.1 Material Specification

's list of specifications for the principal ferritic materials, austenitic stainless steels, bolting and weld materials, which are part of the reactor coolant pressure boundary is given in Table 5.2-2.

Studies have shown that the irradiation induced mechanical property changes of SA-533B materials can depend significantly upon the amount of residual alements present in the compositions, namely; copper, phosphorous, and vanadium. It has also been found that residual sulfur affects the initial toughness of SA-5338 materials. Specific controls are placed on the residual chemistry of reactor vessel plates and the as-depositied welds used to join these plates to limit the maximum predicted increase in the reference temperature (RT , which is discussed in Section 5.3.1.6) and to limit the extent of the rEtor vessel beltline. The beltline is defined by Appendix G of 10CFR50.

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designed fcr oither tensilo cr buckling loads. An cff;rt h:s been made to avoid the use of thin plates which may collapse when subjected to differential pressure.

5.4.2.4 Steam Generator Materials The pressure boundary materials used in the construction of the steam generator are listed in Table 5.2-2.

These materials are in accordance with the ASME Boiler and Pressure Vessel Code Section III. Code cases used in the fabrication of the steam generator are discussed in Section 5.2.1.

The Class 1 components of the steam generator will meet the fracture toughness requirements of the in testing is included ASME code.

Section An additional discussion of fracture toughness 5.2.3.

Discussion of the techniques used to maintain cleanliness during final assembly and shipment are discussed in Section 5.2.3. Onsito cleaning and cleanliness control for the steam generator is discussed in the Applicant's SAR.

5.4.2.4.1 Steam Generator Tubes The method of fastening tubes to the tube sheet conforms with the requirements of Section III and IX of the ASME Code. Tube expansion into the tube sheet -

is total with no voids or crevices occurring along the length of the tube in the tube sheet.

Localized corrosion of tubing material has led to steam generator tcbe leakage in some operating reactor plants. Examination of tube defects that have resulted in leskage has shown that two mechanisms are primarily responsi-ble.

These localized corrosion mechanisms are referred to as (1) Stress assisted caustic cracking, and (2) wastage or beavering. Both of these i

l types of corrosion have been related to steam generators that have operated on phosphate chenistry. The caustic stress corrosion type of failure is precluded by contro111pg bulk water chemistry to the specification limits shown in Section 10.3,2. Renoudt of solids from the secondary side of the g

steam generator is discussed in Section 10.4.4.

Localized wastage or beavering has been eliminated by removing phosphates i from the chemistry control program.

Volatile chemistry (discussed in Section 10.3. ) has been successfully used i in all C-E steam generators that have gone into operation since 1972.

5.4.2.5 Tests and Inspections l

I Prior to, during and after fabrication of the steam generator, nondestruc-tive tests based upon Section III of the ASME Code are performed.

5.4-11

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"a level, resulting in a transient pressure below normal operating piessure.

To minimize the extent of this transient, the backup heaters are energizad, contributing more heat to the water. Backup heaters are deenergized in the event of concurrent high-level error and high pressurizer pressure signals. ,

A low-low pressurizer water level signal deenergizes all heaters before they are uncovered to prevent heater damage. The pressure control program l

is shown in Figure 5.4.10-5.

5.4.10.3 . Evaluation It is demonstrated by analysis in accordance with requirements for ASME Code,Section III, Class 1 vessels that the pressurizer is adequate for all normal operating and transient conditions expected during the life of the facility. Following completion of fabrication, the pressurizer is subjected to the required ASME Code,Section III hydrostatic test and post-hydrostatic test non-destructive testing.

During hot functional testing, the transient performance of the pressurizer is checked by determining its normal heat losses and maximum pr:::r-irrtie- l an6depressurizationratt/. This information is used in setting the pressure I controllers.

Further assurance of the structural integrity of the pressurizer during plant life will be obtained from the inservice inspections performed in accordance with ASME Code,Section XI and described in Section 5.2.

Overpressure protection of the Reactor Coolant System is provided by four

ASME Code spring-loaded safety valves. Refer to Section 5.4.12 and 5.4.13.

5.4.10.4 Tests and Inspections Prior to and during fabrication of the pressurizer, non-destructive testing l is performed in accordance with the requirements of Section III of the ASME Boiler and Pressure Vessel Code. Table 5.4.10-2 summarizes the pressurizer inspecticn program, which also includes tests not required by the Code.

Refer to Section 5.2.1 for inservice inspections of the pressurizer.

~. -

4 t

! 5.4-39 l

TABLE 6.3.3.3-1 SAFETY INJECTION PUMPS MINIMUM OELIVERED FLOW TO RCS (Assuming One Emergency Generator Failed)

Flow Rate Per Injection Point, (gam)

RCS Pressure A A 8 8 (psic) 1 2 1 2 1775.0 0 0 0 0 1650.0 50.0 50.0 50.0 50.0 1440.0 100.0 100.0 100.0 100.0 1270.0 125.0 125.0 125.0 125.0 1095.0 150.0 150.0 150.0 150.0 865.0 175.0 175.0 175.0 175.0 605.0 200.0 200.0 200.0 200.0 310.0 225.0 225.0 225.0 225.0 200.0 234.0 234.0 234.0 234.0 plao +W-0 55f.8?'?? 0 Sld!'!0 0 JW*0#-0 *

• N 100.0 12 ele 170;.0#2Sd 700.0 243.0 243.0 50.0 4g34e 2 00. 0 8084d GM(r 0- 246.0 246.0 0 MM.e 96#-98673 25^0. 0 250.0 25,0.0

• Injection Point Al is assumed to be attached to the broken pump discharge leg.

. _ _ . . , _ _ - - . _ _ . . _ . _ . . _ . . , _ _ _,....-_-_,_.-._._..d_. , , _ _ _ . - . . _ . . _ . _ . , . , , _ , _ . . . _ _ . , . . _ _ _ _ , - . _ . . _ - . . , - , _ _ _

TABLE 6.3.3.3-2 1E';E%. SYSTEM PARAMETER A!10 IflITIAL C6ic::: ;S sitALL BREAK ECCS PERFORMA?lCE AtiALYS:5 Quantity 'ta l ue Units Reactor Power Level (102". of flominal) 3876 ftWt Average Linear Heat Rate (102" of Nominal) 5.6 kw/ft Peak Linear Heat date -

15.0 kw/ft Gap Conductance at Peak Linear Heat Rate 1497 btu /hr-ft 'F Fuel Centerline Temperature at Peak Linear Heat Rate 3681 *F Fuel Average Temperature at Peak Linear Heat Rate 2319 'F Hot Rod Gas Pressure 1187 psia ModeratorTemperatureCoef[icientatInitial ~

Density 0.0 ..: /

• F ,

6 System Flow Rate (Total) 164.0x10 lbs/hr 6

Core Flow Rate 159.1x10 lbs/hr Initial System Pressure 2250 psia Core Inlet Temperature 565 'F Core Outlet Temperature 623 'F Low Pressurizer Pressure Scram Setpoint 1600 psia Safety Injection Actuation Signal Setpoint 1600 psia Safety Injection Tank Pressure 608 osia High Pressure Safety Iniggt_fon Pump Shutoff ,

Head 1773 ?sig Low Pressure Safety Injection Pump Shutoff Head M% psi 9

-y- -*---*-----y--e e-e 1 C*----- -m-- * -

c -++ ---- --W s - T--- m- - - - - - -- c--e t-- w--w-y -- *v^+-tem-r

Lpi w;lk c.

foil win $ G yce.

10- , , , , ,

l l

m, -

- 8 -

x M I 50 8a 6 - -

i$ i cd -

=

S 4 -

t3 w

?

C w 2 -

M U

' ' ' ' \

0 20 40 CO 80 100 120 TIME AFTER RUPI'URE, SECONDS

'e-e:-t 1 4 r ; '5, '??'

l

' 1.0 x DOUBLE EilDED GUIL OT NE DREAK Fi re C-E f E SAFETY lilJECT3l1NW ikkflk Ab DISCHARGE LEG I $f' L

10 ' ' i i i I

1 en O 8 - -

x M

!C en 9 6 - _ _ .

E l .

d E _

5 4 -

W

.E:

, C

W 2 -

Os

' ' ' i i 0

0 20 40 60 80 100 120 f

! TIllE AFTER RUPTURE, SECONDS

(

C-E 1.0 x DOI:BLE ENDED GUILLOTINE BREAK Figur.

! IN PUMP DIScilARGE LEG 6,3,3.

! SAFETY INJECTION FL0li INTO INTACT DISCHARGE LEG 2-5L l

l l

7.16.4.2 Provisions shall be made to preclude the introduction of air into the SCST er ef;;f ; ::rt f r: during fill operations. I 7.16.4.3 If : pr ; f: :::d t: tr:nef:r hydr::fn: t: th: SCST, :11 :f th:

7

• g g met: 1 :-^^ rent: c' th: p r; th:21d 5: ty;; 200 :r Ot? :t:f rl:::

St:01. S21: 05:21d 5: th - :h:nt::1 ty;; cr.d :h;;ld 5: ;;r.- tr;; t;d

- - ,u-,. +- +w. ..._- ...-, m. ..

m -- - _ ,_ , __, . .

. __w___ ,_._ u..n..,__

7.16.5 FIRE PROTECTION A fire protection system shall be I,rovided to protect the Iodine Removal System and shall include, as a minimum, the following features:

a. Facilities for fire detection and alanning.

b. Facilities or methods to minimize the probability of fire and its associated effects.

c. Facilities for fire extinguishment.

d. Methods of fire prevention such as use of fire resistant and non-combustible materials whenever practical, and minimizing exposure of combustible materials to fire hazards.

e. Assurance that fire protection systems do not adversely affect the functional and structural integrity of I safety related structures, systems, and components.

f. Care should be exercised to ensure fire protection systems are designed to assure that their rupture or inadvertent operation does not significantly impair the capability of safety related structures, systems, and components.

7.17 ENVIRONMENTAL 1

1 See Section 7.7 and CESSAR Section 3.11 for environmental interfaces.

7.18 MECHANICAL INTERACTION 7.18.1 IRS compor.ents shall be properly supported such that pipe stresses and support reactions are within allowable limits, as defined in CESSAR Section 3.9.2. CE provides the Applicant the loads at the supports / structures interface locations for components that CE l supplies, under normal, upset, emergency, faulted, and test conditions, as described in CESSAR Section 3.8.5.

l l 7.18.2 IRS piping and fittings shall be Seismic Category I.

t 68-26

s.

1M34AT b

7.16.4.3 . All transfer lines and pump components in contact with the hydrazine solution should be clean and hydrazine compatable as recommended by the chemical manufacturers.

i l

i

(

k 4

4 '

t i

4 f

+

J d

1.75, " Physical Independence of Electric Systems". A discussion of the ".

physical independence is provided below which describes the compliance with Section 4.6 of IEEE 279-1971 and General Design Criteria 3 and 21. General Design Criterion 17 is discussed in the Applicant's Safety Analysis Report.

The PPS cabinet is divided into four bays which are separated by mechanical and thermal barriers. Each bay contains one of the four redundant channels of the RPS and ESFAS. This provides the separation and independence necessary to meet the. requirements of Section 4.6 of IEEE 279-1971.

M M8I _ _ _ _ _ _ i . . _ _ _ a i , u.___,____..- m____. ... .u.. __ _,__i_ ___ m ,_

Z Z' E 'Z E' ~uz .;'.I'un Z . C R.~ :_"' '"- "' ' _ Z ' ' ..'.' ' 'K. "'

iZ:

. _. Z i ..Z,- T i iVZC T'Z GE_l i.~._'1 !!' Z:'"' . ' ".'E '

T.C i,.';4 Z ';'," L "L.a Z _'. '<;'.'i' G ii!.iZ I'"!E T4 Zi "_si' _. .

,Z. W. i. C., ".". .,.C 4. iV 'C. ;l_Li. s.L..L.' <. . . I. .". i. m~

v. 'Z'< . . . ;"i _, . ' i. i. .i.C. . . -" ' ~

The ESFAS Auxiliary Relay Cabinets provide separation and independence for the selective two-out-of-four actuation logics and actuation relays of the two redundant ESF Systems' Trains. Each train's logic and relays are contained in a separate cabinet with all of the train A actuation circuits in one cabinet and all of the train B actuation circuits in the other cabinet. There are mechanical and thermal barriers within the cabinets to protect different portions of the selective two-out-of-four logic from spurious actuation. The two cabinets are physically separated from each other.

T-The RTSS consists of four RTSG. Each RTSG and its associated switches, contacts, relays, etc. is contained in a separate cabinet. Each cabinet is l physically separated from the other cabinets. This method of construction ensures that a single credible failure in one RTSG cannot cause malfunction

or failure in another cabinet.

The separation and independence of the power supplies for each of the above systems is discussed in Chapter 8.0. The interface requirements appear in Section 7.1.3 while ttia implementation will appear in the Applicant's Safety Analysis Report. Protection system analog signals, sent to the Plant Monitoring System (PMS), are isolated from the protection system.

Digital signals are also isolated for the associated signals coming from the protection system.

All of these isolation techniques ensure that no credible failures on the output side of the isolation device will effect the PPS side and that the independence of the PPS is not jeopardized. The test results reports on the isolation devices (within CESSAR Licensing scope) will be submitted for review prior to installation of the devices in the first Applicant's facility.

7.1.2.11 Conformance to IEEE 387-1972 i Conformance to IEEE 387-1972 , "IEEE Trial-Use Standard: Criteria for Diesel-Generator Units Applied as Standby Power Supplies for Nuclear Power Generating Stations", as criteria in the design of these systems is discussed in the Appifcant's Safety Analysis Report.

7.1-8

CE3SAR 7.1.2.10 Replace paragraph 3 with the following paragraph.

Separation of redundant Class IE circuits within the PPS cabinet is accom-1 p11shed through 6 inch separation or barriers or conduit. However, in the for-mation of the logic matrices (AB, AC, BC, AD, 80, CD), initiation circuits, and

- actuation circuits, 6 inch separation is not maintained, nor can barriers or i conduit be utilized. An analysis has been performed to show that the separa-tion achieved is acceptable. Tests and analyses have also been completed to demonstrate that no single credible event in one PPS bay can prevent the circui-try in any other bay from performing its safety function.

i

't v--++ewy----mv4 W-r ""'"'C"" " * ' * " * **'**"~"- -

e-TABLE 9.2-1 -

-96Hel&ANIMEGGNGMl4-s% Tem btMwttAu MA E FF t.utNT MAKEUP WATERALIMITS pH* 6.0 to 8.0 Conductivity Less than 4 pahos O.005 Chloride Less than h ppm Cl Fluoride Less than pm F S :;^-if S & '_ : : tE:n O.5 ;;

i

1. 5m.. . . 19 kl _ _ _ R _ _ _ _ A J / R _ _ _ _ _ A _ J l a ss tha 1pm Sodium Len than 0.003 gm

• If water contains CO , the pH specification may be lowered to 5.8 to compensate for CO absorption.

2

• D:::r:ti:r gi;:: :::::r::ti:: t: :E::p .;:t:r :y:t:: d::i;;n but it i:

net :: :i tr:d ::::::ry.

l zur n:rtrin: t: :::: e ry 7:E:r; .;;t:r 173 i

i

time when exposed to normal reactor coolant chemistry conditions, approaching low steady state values within approximately 200 days. The high pH condition produced by high ammonia concentration (to 50 ppm) minimizes corrosion product release and assists in the rapid development of the passive oxide film. Most of the film is established within 7 days at hot, high pH conditions.

To aid in maintaining the pH dcring this passivation period, lithium in the form of lithiure hydroxide, is added to the coolant and maintained within a 1-2 ppm lithium-7 range.

< At powar, oxygen concentration is limited by maintaining excess dissolved l hydrogen gas in the coolant. The excess hydrogen forces the water 1 decomposition / synthesis reaction in the reactor core to water rather than l Oxygen in the makeup water is removed in the same hydrogen and oxygen.

way.

k In order to minimize the effect of crud deposition en the reactor core heat transfer surfaces, lithium-7 hydroxide additions to the reactor coolant are made. The lithium-7 hydroxide produces pH conditions within the reactor i coolant at operating temperature which reduces the corrosion product solu-bility and, hence, the dissolved crud inventory in the circulating reactor coolant. The elevated pH promotes conditions within the coolant for selective  ;

deposition of corrosion products on cooler surfaces (SG) rather than hotter surfaces (core). An additional advantage is the formation of a more stable j and tenacious passive oxide layer on out-of-core syste_m surfaces. The l.0-2.0 -

lithium concentration is maintained within a 0.2 '.0 ppm lithium-7 range during operation.

_.1_.. _,__ ,,.u._ L:t: *:ru '_

.. .u. _u _' f , '--

2 _, .u. b<__.2 r:: :n- ntr:tim_r m<-

- -.., 2 ,- > i: 'n,

,---m-

_2 i&L.- _*a L- 6 , f**_ au,. __-.__ .L-_ AL- JL--af-- f-_ - L--- -

E _; T_',il'7 ';___:'_:;;_; WI__"'_ II'I';7:27"_':1_':ZTTi__

4 C - ' ! , L _ _ ] ' _ ; i h ! I i L ' I'.'I' ' n T n T _

T I V ' I " '

' 77; G.T V ' m- '

l ---

'A[,5EE5'iE'n: # '-7:EhEEr:t'Ehbr'5bEIO:U.[

9.3.4.1.3.3 Reactivity Control. Boron concentratica is normally I controlled by feed and bleed. To change concentration, the makeup system supplies either reactor maket.p water or boric acid to the Volume Control Tank, and the letdown stream is diverted to the holdup tanks via the preholdup ion exchanger and the gas stripper. Toward the end of a fuel cycle, with.

low boric acid concentration in the coolant, feed and bleed becomes ineffi-cient, and the deborating fon exchanger is used to reduce the RCS boron concentration. The ion exchanger contains an anion resin initially in the hydroxyl form, which is converted to a borate form as boron is removed from the reactor coolant.

9.3.4.2 System Description i

9.3.4.2.1 System .

The normal reactor coolant flow path through the CVCS is indicated by the heavy lines on the Piping and Instrumentation Diagrams, Figures 9.3-1 through 9.3-4.

9.3-5

TABLE NO. 9.3-1 1

(Sheet 1 of 2)

OPERATING LIMITS M AC. TOR Coot.,AMT 1.0 MAKEUP WATER Analysis Normal ?t : :? -

Chloride (C1) <0.15 ppm .15 ;;r S!'f:: (S!0g) '0.01 ;;:  : .02 ;;r C: $ ct?"fty 0 p-'

'c: '

'2 ;:5.::/::

pH 6.0 - 8.0 (1) -

Fluoride (F) 44)- <O.1 ppe 'O.'  ;;r Suspended Solids <0.5 ppm -

2.0 PRIMARY WATER 1s'.he.\

h Gort Core load and Analysis Hot Functionals 9)(7.) !riti ? Criticality Fx: Operation pH (77*) 9.0 - 10.4 4.5 - 10.2N 4.5 - 10. N Conductivity -(4-)-(3) W (~5) -(+F (.3)

Hydrazine 30 - 50 ppm 30 - 50 ppm 1.5 ,: Oxygen ppm (5)

(max. 20 ppm) l

Ammonia <50 ppm <50 ppm <0.5 ppm 1

Dissolved Gas --- ---

(6) 1.0 - 2.0 1. O - 2.0 g Lithium 1-2 ppe 0.2 '.0 ppm 4 0.2 ' 0 ppm 25" i

l Hydrogen ---

---(,/) -

50 cc (STP)/kg (H2 O) (S) 0xygen 40.1 ppe .g0.1 ppm M 10) 50.1 ppm (.t 0I l

Suspended Solids <0.5 ppm, 2 ppm max <0.5 ppm, 2 ppm max <0.5 ppm,2ppmmah Chloride 40.15 ppm 40.15 ppm 40.15 ppm Fluoride 40.1 ppm .g0.1 ppm 40.1 ppm i

l Boron ---

< Refueling '"00 ;;r j Concentration g gg g; eese,.ae1Vm l

. . ~ - - - , ,-

TABLE 9.3-1 (Cont'd)(Sheet 2of2)

Notes for Table No. 9.3-1 j (1) May be as low as 5.8 ff proven due to CO2 absorption.  ;

(2) c! r 'd: '*-ft  ;;!'- d!: t: =t:r '-trd:d f:r pr'-- y rd=;.

(h -fa.). ' Special hot conditioning limits:

Temperature >350'F for 7-10 days

(') O r' ; = = ! = d = d ref=!' ;, ;"'=y 5: n !:e = ? 9 d= t: 'f;h h== =:= txtf r (3) Consistent with additive concentration.

7 astat

( ) Prior to a depressurization shutdown, reduce total gas to <10cc(STP g (H2 O) to limit the possibility for explosive mixtures.

T wstAT h., t< tk p -4<< cat 5 <e- exc51.;e- te 5 ' ; --t= .ted ;<. th 7 L. : ' 'n, 7

ri-t:f '-2 ;;r Li = ti' = te n tf= ': r= h:d ('nd' :t:d b; L' br=hth=u;h),th= = =rt t: th: 2.2-1 ;;: = r.;; . Th': r= ;; := t t:

• - fere prier t: =fti=1fty. 'h: i f tht =:.;: d=: =t ;;;;3 ir. ':; ::= '=i' ;. {

(S) "=r =d :f f:, ch= th: dd:=t'n; != =:h=;n ': ;? =:d '-

ee vice (~?e ;;-e) the 'tthf = 't i: 0.2-0.5 ;;;.

l (9} " t ;;!fe 9?: f r ':; = = ? = d.

L scar Le

@4) (. M yJ c a st n e. Is me.ida641 .4 30.50 pm e,4 3 A Rc5 is \ e ss R.- \so ":,

j (.5) Peter +o .wce A 63 150* F Jun. he. hap oc be.\.a 400*F Aue:q coo) Joua , 3 Tuseer @

4e. &c adi. Fr.- pod core. 4. operafias 3 (7) he:3 sk. 1) be . ta).iael in A. 15 h 25' ca.(st@/w3(n, j kgJr.geo L e b w ieite Ae3*.ssin3 reguire-eds ;% e se &,e r**g%.

re.u e paA naa % e. skaboa- ask barressacheb . '

l (8) H gAr.y st .\J be e5 ee.(.Mh3(.No) Let..,. se.c ,i.3

& cc.d.e e..\aw& fw q s.

(*)) Tk. A a., ) s.mA :b e e .C o.s b .2.o g. is pec ;HeA for up b 14 kenes b att.w be ec A Larsf e. A,bs.

(to) blok .ppbc Me. & a< 43 e. ce. lo A ,

t

Secondary water chemistry is based on tha zero solids treatment method.

This method employs the use of volatile additives to maintain system pH and to scavenga dissolved oxygen present in the feedwater. i A neutralizing amine is added to establish and maintain alkaline conditions in the feedtrain. Neutralizing amines which can be used for pH control are ammoniz, morpholine, and cyclohexylamine. Amonia should be used in plants employing condensate polishing to avoid resin fouling. Although the amines are volatile and will not concentrate in the steam generator, they will reach an equilibrium level which will establi h an alkaline cordition in the steam generator.

Hydrazine is added to scavenge dissolved oxygen present in the feedwater.

Hydrazine also tends to promote the formation of a protective oxide layer on metal surfaces by keeping these layers in a reduced chemical state.

Both the pH agent and hydrazine can be injected continuously at the discharge headers of the condensate pumps or condensate demineralizer, if installed.

These chemicals are added as necessary for chemistry control, and can also be added to the upper steam generator feed line when necessary.

Operating chemistry limits for ft: cter2ndsecondarysteamgjneratr M

..;tc. are give in Tables 10.3.4-1 and 10.3.4-2. 8* ** s con)

• ensc. 'W The limits sta.ted are divided into two groups; normal and abnormal. The limits provide high quality chemistry control and yet permit operating flexibility. The normal chemistry conditions can be maintained by any plant operating with little or no condenser leakage. The abnormal steam generator limits are suggested to permit operations with minor system fault [

conditions until the affected component can be isolated and/or repaired.

The following procedures are recommended to the applicant:

!' th: :::t' ucu: r:rit:r: :n the :t::r g:: r:t:r ble.;da; :::pl: 1'n::

d;t;:t mer th:r four arh::/:r spe '#i: conducti'!*ty, th: :t ;r 3:ncr;t:r c::ter th:uld 5e

• ediately :: ?!cd 2nd 2n:1 ped f:r chl:rf de ::n; ntr:tf;r-

?. If chlorid: ::ncentr:ticr i ithi- 5:::l'n * >: luer, Oper:ttent r:y

ntinue. T' g:ncr:t:r: :h:uld b; :;rpled f:r chi ri#.: :t 1;;;t On 0 pc e ;St "Our thi#t "l: ::nductf;fty encred; f ur er h::/cr i

S. If chl:rtd: ::ncentrati:n i; i On:::: Of 5:: l'n * ;;lue: Or :ppr;;;h'n; 0.1 pre, : nd::::r 10:E i: *ndt::ted 2nd ' :E 1:012tfer precedure:

th:uld be 'nttituted.

'S:: lin: ; ! :: re defi ed n:: the chltrid: 00ncentr:t!O- ! S nfet: 4-th: :t::: ; ncr:t:r .;;t:r :t :t::dy :t:t; :per; ting ::nditi:n; iht :st 0:nd:ncer 10:h:;0. ^ typic:1 5::elin: c'!:rfde cencentr:tter i: enpected u <_ ____,__m .m __<

ww ma w wyyu vn..-ww aj mv 77-.

N 10.3-2

s 1-,@ .

When the normal range is exceeded, inmediate investigation of'the problem should be initiated, sampling frequency increased to the abnormal level (at least twice per 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shift) and blowdown increased to one (1) percent of the main steaming rate. The problem shculd be corrected and the parameter (s) returned to the normal range within one week. If this cannot be done, and the parameter has a listed abnormal range, power should be reduced to 25%

as if the abnormal range had been exceeded.

When the abnormal range is exceeded, power should be reduced to the lowest value (maximum of 25%) consistent with automatic operation of the feed system. Continued plant operation is then possible while corrective action is taken. Power reduction should be initiated within four hours of exceeding the abnormal range. The p*oblem should be corrected and the parameter (s) returned to 'he normal range within one hundred (100) hours. If this cannot be done, the unit should be shutdown.

L _ _ . _ _ _ - _ _ _

l If ry Of the c r2! 0; r:ttr; :; rf#ft:tt::: '- T d k 10.0.0 1 ;c; ;,,;;;d;d ,

'nrd* te 4 fert'; t'e e' the pr:5! r 2::!d 5: i-i.ti:t;d , =:.,.'. i;;; f7;;;=;;

est- te th- der:! 10fe! ( t le=t tt:f r pr :f ght h:r :hift) =d j bit:dt: *nr: red ' ::: ;; rent Of the ri ster'n; r:te. If:=d=:r  !

led ;c it i-dfested, E d t r h tfe precedur= d =1d 5: ' :titut:d. ";r

r"t=ttr Of a

=d==r 1:2. "tth ster ;rr:tr durid: in===

Of 0.? ;; , phat ;; cr r:d=ti= deeld 50 i-itt:ted ix;di tcly =d o

c'^uld be reduced t 25! "fth4- f= r he re. n:r d=1d n t =:=d 25%

=tt' chkrid: :==ntr:thn h "ith*- the ner ' :pe:f fictt =.

!r the ev=t :=d:=:r hd ;c === ry 0;r: ting ;=iff =ti = f r Tdi:

70.3.t i t: ;= ? Or :==d d=r:1 -it:, : detdn: :h =id 5: :=;1;ted "ith'- f:r 5: r;. Draining or flushing of the steam generators will be necessary to reduce the impurity concentration.

10.3.4.2 Corrosion Control Effectiveness Alkaline conditions in the feedtrain and the steam generator reduce general corrosion at elevated temperatures and tend to decrease the release of soluble corrosion products from metal surfaces. These conditions promote the formation of a protective metal oxide film and thus reduce the corrosion products released into the steam generator.

Hydrazine also promotes the formation of a metal oxide film by the reduction of ferric oxide to magnetite. Ferric oxide may be loosened from the metal surfaces and be transported by the feedwater. Magnetite, however, provides' an adherent protective layer on carbon steel surfaces. Hydrazine also promotes the formation of protective metal oxide layers on copper surfaces.

The removal of oxygen from the secondary waters is also essential in reducing corrosion. Oxygen dissolved in water causes general corrosion that can result in pitting of ferrous metals, particularly carbon steel. Oxygen is removed from the steam cycle condensate in the main condenser deaerating section. Additional oxygen prctection is obtained by chemical injection of hydrazine into the condensate stream. Maintaining a residual level of hydrazine in the feedwater ensures that any dissolved oxygen not removed by the main condenser is scavenged before it can enter the steam generator.

The presence of free hydroxide (OH-) can cause rapid corrosion (caustic stress corrosion) if it is allowed to concentrate in a local area. Free hydroxide is avoided by maintaining proper pH control, and by minimizing impurity ingress in the steam generator.

10.3-2(a)

TABLE 10.3.4-1 1

OPERATING CHEMISTRY LIMITS FOR l SECONDARY STEAM GENERATOR WATER Normal (I) Abnormal N

' Variable Slecifications Limits a

pH ( m ued sgstem -

9.2 -7. 5 .2; j l

(, ew.c 6.) 90-9.4 se ,

9.,

cA.a Conductivity (3) oa

< A umhos/cm.g.).

o.s-2.o 15 pmhos/cm

";;:i'f: -7 S r ; ded Sc!4dr ' 0 pr

!0 pp-Silica 9 90. n.tt;h y m--

10 ;;r Chloride 9' 0 6;;r 2.0 - 100 ppL (4')

S. Atua e 20 rek 20 - 100 ppt Sui 9ata 4 15 ppk I 5 -100 ppb NOTES:

l (1) Normal specifications are those which should be maintained by continuous l

steam generator blowdown during proper operation of secondary systems.

'2' "

..t: . nl -it: 'nd' :t^ : ': !! :^nditi r : ict: : d ^?: t :h: tdt:-

chee!d 5: e .::d << ere 2! '<-it: : e c= :cded r e- ' 'cu .

r,, e___sn:. . . >. . s s . n s , . ,, s a g ., , ,y .. . ....__... ..,---- r--n.s- _ ',::l:-

l (2) M ix. A syne- is a3 su..aAa 3 3skam coda'ia la$ .caff*'

a\\on caneoseds,

! (3) 19- 4Le i- - =A *ke SLddoua li~;f *S ?0/*k'5/cm is av.eu).A &a ua;k .skoJ A be SkdAod" 'dikkia t

cou,e kenes.

[ @ T9 A immeAide. skJA e 1:m;k .0 500 pg is a.c skeJ 4ke u:\- .sk.J A \. < ' sLJJoce w ; W. .,,

C. < kowes.

l

-.--,w .,---- .-- .,n re-~.,-- ,,...-,,-,,,,,,,-,,,-,,-,.m ,,,e,,

,.,,,,.ng,,wnm,,

.,,m-.w.,,,,,.,,., .,-.,w,,w,-,,,,,-,~,,.-,y.,m-y,_,,q.--p,

TABLE 10.3.4-2 -

Foll.

OPERATING CHEMISTRY LIMITSAFEEDWATER AND couMMsaT1r Normal (I) ft=r2 O Variable Specifications Li-it; pH mixed

a. -Fe i:t:r system =nt 'nin;

p;cr :!!:y : ;:n:nt: 8.8 - 9.2

b. Copper-free f= i::t:r 9.,.g - g, g system . ...

Conductivity (Intensified 0.2 cation) (F adw.t.c) < 9-4 umhos/cm 0.5 ' .5 ,.rh = ,' r Hydrazine (FeaAwak*') 10 - 50 ppb h selved A 0xy en 3 Feed) <+ ppb (2) 5 IO ??I Condensate) < 10 ppb 10 ;;5

_, Sodium <hppb '

10 ;;b Copper (.Fa*Aud*") -fat- e.2, (2)

-Iron (Fe Jwahc) <Sppb 10 20 ;;5 1- %.

(bith =pp;r) ' 0 ;p" -

(..ith=t =;;;r) 2.0 ;;r

= , :: ;;t yH c.Jr.\ AAAttiv. (5)

NOTES:

(1) Normal specifications are those which should be maintained during proper operation of secondary systems.

(2) ?i = r ! 't: ' nd' n t: : f=lt =nditix si;t: =d pl;c.t Athn J.; M 5: 1 = =d e = rci -it: : = = := d:d fx ' h= c:.

(2) El ' '=t: =;; r 'n; = = te th: :t= ;=:= t:= t:= pr=tini .

(jit) Tbs e.a Asa s de. b a.c - .\ \t-ik is t o -30 pek bo.4 &ba

<e ptrament C.e imme A +t skJ A.wn Aces at nig even ik 4ke ecokiem is ad e.ereded wikk6 \00 bcs.

(3) F.e +ka a.ad ensde. 3 sodium is ~. d oceA d e e.b.

conAeaste Let we.t1

(.4) Aa \ $s s nd capired for copper Gese. 5 3s4**s.

(S) \ im* is A.p.aJed mean pH,

9 r 2. 0 _ PREREQUISITES

~

\

2.1 been calibrated. Pressurizer pressure and level control system instru 2.2 i Support systems required for the operation of the pressurizer

-r pressure and level control systems are operational.

2.3 Test equipment is available and calibrated.

3.0 TEST METHOD 3.1

) Simulate a decreasing pressurizer pressure and observe heater response and alarm and interlock setpoints.

} 3.2

Simulato an increasing pressurizer pressure and observe heater and spray valve response and alarm and interlock setpoints.

3.3 charging pump response and alarm and interlock s 3.4 charging pump response and alarm and interlock s 3.5 Simulate a low pressurizer level and observe operation of the 3.6 letdownCouTAcu ieW"- - valve $S Simulate and alarm and a low-low interlockpressurizer setpoints. level and obsarve heater response 4.0 DATA REQUIRED 4.1 Response of pressurizer heaters to simulated pressure and level signals.

, 4.2 Response of spray valves to simulated pressurizer pressure.

4.3 Response of charging pumps to simulated pressurizer level.

4.4 CON 1M-Response of letdownkalve#Eo simulated pressurizer level.

4.5 Cowtgoc. r-S l Response of letdown ica1= +

level. 4= valve 9 o simulated low pressurizer ,

4.6 I

Values of parameters at which alarms and interlocks occur.

I l

l 14.2-59

" w -- , -,

--"e-- n s , , em ,-m ,

w p- ,r e-- w wee -or-- - - - - -- w,-,-e- ,-

---me- , , - - < -

HOT SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 4.1. 3 At least two of the loop (s)/ train (s) listed below shall be OPERA 8LE and at least one Reactor Coolant and/or shutdown cooling loop shall be in operation.*

a. Reactor Coolant Loop #A and its associated steam generator and at least one associated Reactor Coolant pump,**

l

b. Reactor Coolant Loop #8 and its associated steam generator and at i least one associated Reactor Coolant pump,**

c. Shutdown Cooling Train #1,

d. Shutdown Cooling Train #2.

! APPLICA81LITY: MODE 4 (T Cold 1 3M*F)

ACT[0,N,,: ,

a. With less than the above required Reactor Coolant and/or shutdown cooling loops OPERA 8LE, immediately initiate corrective action to .

return the required loops to OPERA 8LE status as soon as possible; if the remaining OPERA 8LE loop is a shutdown cooiing loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With no Reactor Coolant or shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration l of the Reactor Coolant System and innediately initiate corrective action to retum the required coolant loop to operation, i SURVEILLANCE REQUIREMENTS 4.4.1.3.1 The required Reactor Coolant pump (s), if not in operation, shall be

~ determined to be OPERABLE once per 7 days by verifying correct breaker align-ments and indicated power availability.

\

l

'

• All Reactor Coolant pumps and shutdown cooling pumps may be de-energized t for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) cora outlet temperature is maintained at least 10*F below saturation temperature.

• • A Reactor Coolant pump shall not be r. tarted with one or more of the Reactor

(***)*F during Coolant cooldown System or (b)cold**)*F leg during

(* temperatures less heatup unless the than or equal3 to (a) secondary water temperature of each steam generator is less than F above each l l

of the Reactor Coolant System cold leg temperatures.

• • • See, Applicant's SAR.

. 3/4 4-3

COLD SHUTDOWI - LOOPS FILLED LIMITING CON 0! TION FOR OPERATION 3.4.1.4.1 At least one shutdown cooling loop shall be OPERA 8LE'and in opera-tion *, and either:

a. One additional shutdown cooling loop shall be OPERA 8 lei, or

b. The secondary side water level of at least two steam generators 1

shall be greater than 255 on the wide range level indicator.

APPLICABILITY: MODE 5 with Reactor Coolant loops filledH.

ACTION:

a. With less than the above required loops OPERA 8LE or with less than the required steam generator level, inmediately initiate corrective action to return the required loops to OPERA 8LE status or to restore the required level as soon as possible.

b. With no shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and inanediately initiate corrective action to return the

-required shutdown cooling loop to operation.

SURVEILLANCE REQUIREMENTS s 4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least or;ce per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.4.1.2 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at a flowrate greater than or equal to 4000 gpm at least onca per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1. One shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveil-lance testing provided the other shutdown cooling loop is OPERABLE and in operation. ,

N A reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to (**)*F during cooldown or (**)*F during heatup unless the secondary water temperature (saturation temperature corresponding to the steam generator pressure) of each Reactor Coolant steam System generator cold is less thang4YF above each of the leg temperatures.

j

• The shutdown cooling pump may be de-energized for up to I hour provided

1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10*F below saturation temperature.

• • See Applicant's SAR.

l l 3/4 4-5 l

4- o .

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR C0OLANT L00pS AND C00LAMT CIRCULATION

, The plant is designed to operate with both reactor coolant loops and associat-ed reactor coolant pumps in operation, and maintain DNBR above 1.22 during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires tha* the plant be in at least HOT STAN08Y within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat remeval

, capability for removing decay heat; however, single failure considerations

! require that two loops be OPERA 8LE.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reacto:-

coolant loop or shutdown cocling loop provides sufficient heat removal capabil-ity for removing decay heat; but single failure considerations required that at least two loops (either shutdown cooling or RCS) be OPERA 8LE. Thus, if the reactor coolant loops are not OPERA 8LE, this specification requires that two

• shutdown cooling loops be OPERABLE.

In MODE 5 with reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling loops be OPERABLE.

Tha operation of one Reactor Coolant Pumo or one shutdown cooling pump pro-j vides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during baron concentration reductions in the Reactor Coolant System. A flowrate of at least 4000 gpa will circulate one equivalent reactor coolant system volume of 12,097 cubic feet in approximately 23 minutes. The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump during MODES 4 and 5 with one or more RCS cold leg less than or equal to * 'F during cooldown or * *F t

during heatup are provided to present RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of

. Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting start-ing of' the RCPs to when the secondary water temperature of each steam genera-i tor is less thanJ584 above each of the RCS cold leg temperatures. l IWF

• 5ee Applicant's 5AR.

I B 3/4 4-1 i-

m. , , w,,n---.--,--n ,~,,,,r-a-,-,v- -,,, ann ,-n, v., w - ,,-a..,_-a----

The limitations imposed on the pressurizer heatup and cooldown rates and spray

water temperature differential are provided to assure that the pretsurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the' ASE Code requirements.

The OPERA 8ILITY of two shutdown cooling suction line relief valves, one ,

located in each shutdown cooling suction line, while meintaining the limita-tions imposed on the RCS heatup and cooldown rates, ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix 6 to 10 CFR Part 50 when ona or more of the RCS cold legs are less than or equal to * 'F during cooldown or * 'F during heatup. Either one of the two SCS suction line relief valves provide adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water tarserature of the steam generator less than or equal to )459 above the RCS cold leg temper-atures or (2) the inadvertent safety injecti m actuation with two HPSI pumps l injecting into a water solid RCS with full c iarging capacity and with letdown isolated. These events are the most limiti g energy and mass addition trans-4 lents, respectively, when the RCS is at low temperatures.

3/4.4.9 STRUCTURAL INTEGRITY p *p-The inspection programs for the safety-related ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained " n acceptable level throughout the life of the plant. To the ble, the inspection program for these components is in compli-ance witn Section XI of the ASME Boiler and Pressure Vessel Code.

! <i r

~-

• see Applicant's sAR.

B 3/4 4-6

I Regulatory Guide 1.68.2, Revision 1 INITIAL STARTUP TEST PROGRAM TO DEMONSTRATE REMOTE SHUTDOWN CAPABILITY FOR WATER-COOLED NUCLEAR POWER PLANTS l

SUMARY

/

Regulatory Guide 1.68.2, Rev. 1 describes an initial startup test program acceptable to the NRC staff fer demonstrating hot standby capability and the potential for cold shutdown from outside the control room.

POSITION The CESSAR initial startup test program fully meets the intent of Regulatory Guide 1.68.2.

l i

i l

A-36

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