ML19309B111

Surveillance Requirements Review for Secondary Coolant Sys, Revision 1

ML19309B111
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/25/1980
From:	PUBLIC SERVICE CO. OF COLORADO
To:
Shared Package
ML19309B107	List: ML19309B106 ML19309B108 ML19309B109 ML19309B110 ML19309B111 ML19309B112
References
EE-22-0002, EE-22-2, P-80064, TAC-65991, NUDOCS 8004030134
Download: ML19309B111 (28)

Text

- . _ - _ __ . . _ . .-

. .w, Enclosure to p 80064 (4)

FORT ST. VRAIN INSERVICE INSPECTION AND TESTING PROGRAM J

SURVEILLANCE REQUIREMENTS REVIEW FOR THE i

3 SECONDARY COOLANT SYSTEM (22) i l

00<1000})[r i

EE-22-0002 Rev. 1

. $nggh 2 5, 198.0

.-s EE-22-0002 nev, 1 p.1 REPORT EE-22-0002 FORT ST. VRAIN INSERVICE INSPECTION AND TESTING PROGRAM SURVEILLANCE REQUIREMENTS REVIEW FOR THE SECONDARY COOLANT SYSTEM

1. INTRODUCTION A review was performed of the current. surveillance require-ments for the secondary coolant system. As a result of this review, additional or modified surveillance requirements may have been recommended to meet the criteria established for the Fort St. Vrain inservice inspection and testing program which has been presented to the Nuclear Regulatory Commission and which are specified in Ref. 1. The review included the reference do-cuments listed in section 5 of this report, and in particular the proposed ASME Code Section XI, Division 2, Draft. An explanation is provided when the current and recommended sur-veillance differ from the proposed Code requirements. The review also included the operating experience with the equipment at the plant.

This report covers the equipment items shown on system diagrams PI-22-1 through -9, with the exception of the fire water booster pumps (P-2109 and P-2110) and associated system portions (downstream of valves V211565, V22370 and V22371) which will be treated in the report reviewing the surveillance re-quirements for the helium circulator auxiliary system (system 21).

Equipment items shown on PI-22-10 will be reviewed with the turbine steam system (system 52), since they have similar functions.

Also , included in this report is the review of the safety class 1 part of the emergency feedwater and condensate headers shown on system diagram PI-31-1, and the review of the helium circulator drive steam and feedwater ducting.

Excluded from the review are the instrumentation and controls associated with plant protective and overall plant control functions, as well as the radiation monitoring instru-ments, which will be reviewed with the plant instrumentation and control system (system 93). The pipe supports and restraints are also excluded from this review, since they will all be covered under a separate report.

u .2N EE-22-0002 Rev. 1 p.2

2. SURVEILLANCE CLASSIFICATION 2.1 SYSTEM FUNCTIONS In addition to their normal function of generating high pressure main steam and medium pressure reheat steam to drive the main turbine generator and the helium circulators, the steam generators and the secondary coolant system, or parts thereof, have several safety functions as described below.

(a) Safe shutdown core cooling: Except for the case of permanent loss of forced circulation accident, the steam generators are used for residual heat removal from the reactor under normal conditions and under all postulated accident conditions, including the worst environmental conditions where all non safety class I equipment is presumed to have failed. The parts of the system used for safe shutdown cooling are:

- the emergency feedwater header and the emergency con-densate header either of which can supply emergency cooling water to the steam generators and emergency drive water to the helium circulator Pelton wheels,

- the steam generator superheater section and the main feedwater and main steam pipe included between the isolation valves,

- the main steam bypass valves,

- the steam generator reheater section and the cold and hot reheat pipe included between the circulator steam turbine trip valves, circulator bypass isolation valves, and the hot reheat steam stop-check valves,

- the' reheater discharge bypass valves.

(b) Steam / water dump: Following a moisture ingress accident, one loop is tripped, and the steam generator superheater section is isolated and dumped to limit the amount of moisture ingress in the reactor coolant system. Should the wrong loop have been dumped, the pressure in the primary system will continue to increase, and upon reaching the high primary pressure setpoint will cause the leaking loop still in operation to be depressurized, thus providing continuous residual heat removal and  :

limiting the rate of moisture ingress in the reactor coolant system while the intact loop is being brought back to operation. The parts of the secondary coolant system which participate in the steam / water dump function are:

- the steam water dump valves and tank,

- the main steam bypass valves,

- the valves which are used for a loop trip.

. Q EE-22-0002 Rev. 1 p.3 2.1 (cont.)

(c) Loop trip and circulator trip: Valves which are required to close automatically in case of loop trip or circulator trip. This includes valves required to operate for safe shutdown cooling or steam water dump, and additional valves used particularly for circulator trip.

(d) Reactor coolant boundary and containment: The steam generator primary assemblies function as primary contain-ment since they form parts of the reactor coolant boundary.

Parts of the secondary coolant piping.and valves function as secondary containment follcwing a leak or a rupture in the steam generator primary assembly. The main steam and reheat steam safety valves function to protect the integrity of the steam generators and secondary coolant piping and valves.

2.2 SURVEILLANCE CLASSIFICATION Based upon their functions, as described above, the com-ponents of the secondary coolant system have been assigned to surveillance classes as outlined hereafter.

a) Active equipment items which, amongst other functions, participate in safe shutdown cooling, and/or steam / water dump, and/or loop or circulator trip are assigned to surveillance class S2 according to criteria 2.lb, 2.2b and 2.3a of Ref. 1.

b) Active and passive equipment items, which have a contain-ment or containment protection function only, are assigned 4

to surveillance class S3 according to criteria 2.lc, 2.2c and 2.3b of Ref. 1.

c) Other non safety class I equipment items which are used to contain radioactive fluids, or to monitor or control im-portant system components, are assigned to surveillance class S4 according to criteria 2.ld, 2.2d and 2.3c of Ref. 1.

2.3 APPLICABLE SURVEILLANCE CRITERIA The following criteria of Ref. 1 are considered applicable when making the surveillance review of the secondary coolant system.

a) The operational readiness of the parts of the system assigned to surveillance class S2 is to be demonstrated by normal operation or by system testing at a frequency consistent with the highest active component test frequency, to demonstrate system availability and operability to perform its safety function (criteria 3.lb) .

EE-22-0002 Rev. 1 P4 2.3 (cont.)

b) The operational readiness of pumps assigned to surveillance i

class S4 (the only one in the system) is to be demonstrated by normal operation, or by surveillance testing once each quarter to exercise those pumps not normally in operation (criteria 3. 2. lc) .

c) The operational readiness of safety / relief valves assigned to surveillance class S2, S3 or S4 is to be demonstrated by testing at a frequency and according to rules based on the pro-posed Draft ASME Code,Section XI, Division 2 (hereafter referred to as the Code). Operational rea'diness of automatic isolation valves is to be demonstrated by normal operation, j or by surveillance testing at Code frequency for surveillance class S2 or at least once each year for surveillance class S3 and S4, to exercise those valves which do not normally operate according to Code rules and requirements. The same criteria apply to check valves and remote manual isolation valves,

, except that the test frequency is at least once each year for surveillance class S2 valves, and at least once every five years for surveillance class S3 and S4 valves. Criteria 3.2.2d allows that surveillance class S2 valves, which do not normally operate, be tested to the requirements of surveillance class S3 if, during reactor operation, they are in the position re-quired to perform their safety function. Criteria 3.2.2d also exempts from testing the valves which are normally in the po-sition required to perform their safety function and which are fail safe in that position, and the valves exempt from testing by the proposed Code.

d) Operational readiness of instrumentation and controls

, (criteria 3. 2. 3b and 3. 2. 3c) is to be demonstrated by a func-tional test at least quarterly, for surveillance class S2 instrumentation which monitors an active safety function, or at least annually for the remaining surveillance class S2 , S3 or S4 instrumentation. Tests of valve instrumentation and controls is to be performed at the frequency required for valve testing (criteria 3.2.3d). In addition to the above functional tests, instrument accuracy is to be demonstrated generally once each year by a calibration test (criteria 3.2.3e).

e) Structural integrity of safety related passive plant equipment can be demonstrated by continuous leakage monitoring and/or alarm during plant operation (criteria 3.3.la) or, if no leakage monitoring instruments are provided, by an examination for leakage of accessible equipment portions while the equipment is at or near its nor-mal operating pressure (criteria 3.2.3b). Such an examination for leakage is to be performed at least once during each refueling cycle for surveillance class S3 equipment, or at least once every five years for surveillance class S4 equipment.

Where specific concerns have been identified, non destructive i examination may be used to suppLament surveillance inspections and tests. Criteria 3.3.lc exempts from leakage examination

. o EE-22-0002 Rev. 1 1 -

p.5 2.3 (cont.)

those equipment items which operate at condi '.ons not expected to degrade their integrity during normal operation when compared to design conditions, and those equioment items the failure of which does not prevent the performance of an active safety function and does not lead to unacceptable release of radioactivity. Critoria 3.3.2c specifies that the structural integrity of the steam generator reactor coolant boundary is to be verified by continuous leakage monitoring and deemed acceptable as long as primary coolant moisture and reheat steam radioactivity remain below acceptable limits. The structural integrity of the steam generators is further assured by controlling feedwater chemistry.

3. OPERATIONAL READINESS 3.1 OPERATIONAL READINESS OF THE SYSTEM Most parts of the secondary coolant system are used for normal reactor operation at power, for startup and shutdown,

. and during shutdown. The only active equipment items which are not used during normal operation are valves and the dump tank pump. Testing the operational readiness of these valves and the pump is sufficient to demonstrate the operational readiness of the system.

Further, the operational readiness of the system to per-form its safety functions is also assured by compliance with the requirements of technical specifications LCO 4.3.1 and LCO 4.3.3. LCO 4.3.1 requires that at least one steam gen-erator and.either section of the second one be operable for decay heat removal, the emergency feedwater header be capable of supplying water to the operable suoerheater sections, and the emergency condensate header be capable of supplying water to either the operable superheater or reheater sections.

LCO 4.3.3 requires that an appropriate condensate inventory be contained in the steam / water dump tank.

3.2 OPERATIONAL READINESS OF PUMPS a) Current surveillance requirements:

1 The dump tank drain pump (the only pump in the scope of this report) and its stop valve (IIV-2 27 9 ) are operated to perform the functional test of the ~ team / water tank level  ;

, indicators in accordance with technical se ccification SR 5.3.1.

I

. ?

.s .

EE-22-0002 Rev. 1 P.6 3.2 (cont.)

b) Recommended surveillance requirements:

The dump tank drain pump (P-2201) is assigned to sur-veillance class S4, since it may contain radioactive fluids and since its active function is not directly related with a safety function. The test frequency of SR 5.3.1 (see paragraph 3.4.1 hereafter) meets criteria 3.2.lc of Ref. 1. No additional surveillance is recommended since the performance of the pump has no impact on plant safety.

c) Proposed ASME Code requirements:

The dump tank drain pump is not designed according to any Code class. Therefore, subsection IGP does not apply.

d) The recommended surveillance exceeds the proposed Code requirements.

3.3 OPERATIONAL READINESS OF VALVES 3.3.1 AUTOMATIC AND REMOTE MANUAL' ISOLATION VALVES a) Current surveillance requirements:

Technical specification SR 5.3.1 requires that the steam /

water dump valves be tested individually every three months.

Technical specification SR 5.3.2 regrires that the main steam and hot reheat steam stop check valves be full stroke tested once per year with the reactor shutdown and partial stroke tested onc,e per week. Technical specification SR 5.3.3 re-quires that the main steam (and hot reheat steam - system 52) power operated relief valves, the main steam bypass valves (and the six hot reheat steam bypass valves - system 52) be tested once per year during plant shutdown. Technical specification SR 5.3.4 requires that those valves, which are pneumatically, hydraulically, or electrically operated and are required for actuation of the safe shutdown cooling mode of operatior , be tested twice annually, with an allowable interval between two consecutive tests extending from four to eight months.

b) Recommended surveillance requirements:

b.1) Steam / water dump valves (HV2215 thru HV2218)

Full stroke exercising the steam / water dump valves quarterly meets the proposed Code test frequency (see paragraph (c) hereafter). Testing the valves one at a time assures that the steam / water dump system remains operable. Surveillance procedure SR 5.3.1-0 has been reviewed. It provides for measurement of the valve opening and closing times, and for observation of the valve position indicating lights, which is

% ,F '

EE-22-0002 Rev. 1 P.7 3.3.1 (cont.)

adequate for functionally testing the valve instrumentation.

No additional or modified surveillance requirements are recommended for the steam water dump valves, since the rules for testing meet the proposed Code requirements as specified by criteria 3.2.2b of Ref. 1 for surveillance class S2 auto-matic isolation valves.

b.2 Safe shutdown cooling valves:

(i) TECHNICAL SPECIFICATION INTERFACES .

The review of the technical procedures, corresponding to technical specifications SR 5.3.2, SR 5.3.3 and SR 5.3.4, indicates that there is a duplication of testing for some valv's. Since the testing required by SR 5.3.4 is more rigt.ous, it is recommended that technical specifications SR 5.3.2 and SR 5.3.3 be modified to delete the test require-ments included in SR 5.3.4. This concerns the main steam and hot reheat steam stop-check valves (SR 5.3.2 annual full stroke test) and the main steam bypass valves (SR 5.3.3 annual full stroke test).

(ii) LOOP ISOLATION VALVES (HV2201, HV2202, FV2205, FV2206, HV2223, HV2224, HV2241, HV2242, PV2243, PV2244, HV2245 thru HV2254, SV2105, SV2106, SV2111, SV2112)

Amongst the loop isolation valves, technical specification SR 5.3.2 requires that the main steam and hot reheat steam stop-check valves be partial stroke tested once per week. As indicated in Section 10.5 of the FSAR, no partial stroke testing is required during plant operation at power for the feedwater block valves and the circulator outlet / block valves, because each of these valves has a backup in the form of a i

control va,1ve which is also actuated by the plant protective system. Each circulator bypass valve also has a backup in the form of a control valve which, even though it is not actuated by the plant protective system, will close upon action of its safety class 1 control when the loop is tripped.

Therefore, the main loop isolation fun 7 tion is placed on the control valves which, by function, are part stroke exercised during plant operation. Since it would not be practical to full stroke exercise these valves every three months, the part stroke valve testing is adequate to demoratrate the operational readiness of the loop isolation valves until a scheduled plant shutdown when the valves can be full stroke exercised Full stroke exercising of the loop isolation valves is provided by technical specification SR 5.3.4. A review of surveillance procedure SR 5.3.4-SA shows that a full stroke test is performed for the feedwater block and flow valnes, main steam stop-check valves, circulator outlet block / trip valves, circulator bypass pressure control valves, and hot reheat steam stop-check valves. The circulator steam turbine speed control valves (SV 2105, SV 2106, SV 2111, SV 2112) and the

EE-22-0002 Rev. 1 p.8 3.3.1 (cont.)

circulator bypass valves (HV2241 and HV2242) are automatic loop isolation valves not included in this current surveillance pro-cedure. However, these valves, together with the other circula-tor trip valves, are exercised at each refueling shutdown when testing the circulator trip per technical specification SR 5.4.1 (Table 5.4-3, item 7). The circulator cold reheat block valves (HV 2245 through HV 2249) are remote manual valves, the con-trols of which are not safety related. They are used as backup isolation valves and are provided with a manual override.

There is currently no testing of these valves, nor is any recommended due to their function and design features.

The current test frequency is semi-annual and the test requires that the plant be brought to less than 2 percent reactor power (which for all practical purposes ic similar to a reactor shutdown). Technical specification SR 5.3.4 pro-vides some operational flexibility by allowing the interval between two consecutive tests to vary between 4 and 8 months.

In the case under consideration, the proposed Code frequency (applicable per criteria 3.2.2b of Ref. 1) is at each plant cold shutdown, or at 3 month intervals in case of frequent cold shutdowns. Since the interval between two plant cold shutdowns could be as much as a refueling cygle, it is considered that the current surveillance exceeds the proposed Code requirements and criteria 3.2.2b, and that technical specification SR 5.3.4 could be modified to provide more operational flexibility.

(iii) LOOP BYPASS VALVES (PV2229, PV2230, PV2267, PV2268, HV2292, HV2293, PV22129, PV22130, HV22131, HV22132, PV22153, PV22154, PV22167, PV22168, HRBV-1 thru HRBV-6).

Bypasses on the main steam and hot reheat steam lines provide an alternate flow path for the steam or cooling water.

The startup bypass is used at startup to direct the hot water to the preflash taak. The startup bypass block valves (HV 2293, HV 2292) are required to close for loop isolation and safe shutdown cooling. They are tested according to technical specification SR 5.3.4 and otherwise are used at each reactor shutdown and startuo, which is considered adequate to demon-strate their operational readiness. The main steam and hot reheat steam bypasses are used in conjunction with the main steam and hot reheat steam power operated relief valves which discharge to the atmosphere to provide a flow path for the steam in case of turbine trip, thus preventing opening of the safety valves. The main steam bypass is also used to provide a cooling flow path in case of safe shutdown cooling and to depressurize the operating loop, should this be required upon high primary pressure. The main steam bypass and de- )

pressurization valves (PV2229, PV2230, PV22129, PV22130, PV22153, PV22154) are tested according to technical specification SR 5.3.4.

EE-22-0002 Rev. 1 p.9 3.3.1 (cont.)

Partial stroking of these valves is not practical during reactor operation at power, therefore the current test re-quirements are considered adequate. The main steam power operated relief valves (PV22167, PV22168), the hot reheat steam power operated relief valves (PV5 2 21-1, PV5 221-2 ) , and the hot reheat bypass valves (HRBV-1 through -6) are assigned to surveillance class 54, for which criteria 3.2.2c of Ref.1 requires annual testing. The current operational flexibility meets these criteria.

However, to provide operational flexibility, it is recommended that technical specification SR 5.3.3 be modified to allow plant operation until the next scheduled shutdown if the valves have not been tested during the previous year.

The reheater discharge bypass is used to provide a cooling flow path to the condenser when the reheater is flooded for safe shutdown cooling. The reheater discharge bypass block-valves (HV22131, HV22132) and control valves (PV2267, PV2268) are tested in accordance with technical specification SR 5.3.4, the frequency of which is adequate to meet the annual test frequency of criteria 3.2.2b of Ref. 1 for surveillance class S2 remote manual valves. .

(iv) EMERGENCY FEED AND CONDENSATE VALVES (HV2203, HV2204, HV2237, Hv2238, FV2239, FV2240, HV2290, HV2291, HV31118 thru MV71122, HV31191)

Automatically actuated valves in the emergency feedwater header supply emergency feedwater to the steam generator suoer-heater sections HV2203, HV2204. These valves cannot be tested during reactor operation at power, since their controls would also ac'tuate other valves, such as the feedwater block valves, and since they are not designed for partial stroking. They are tested in accordance with specification SR 5.3.4 which is considered adequate to meet criteria 3.2.2b of Ref. 1 and the proposed Code requirements.

Automatically actuated stop-check valves are provided to isolate the non safety related feedwater system from the safety J related emergency feedwater header in case of safe shutdown I cooling (HV31118, HV31119, HV21120), and to prevent loss of feed- l water to the steam generators in case of rupture of the emergency feedwater header. The latter function is not safety related, so that the valve actuators are not safety related.

However, since it prevents an abnormal situation, it is assigned to surveillance class 54 for which the criteria of I Ref. 1 require an annual testing. The criteria are met by l normal plant operation since the valves are cperated during I each plant shutdown. Therefore, the valves can be exempt from testing. The check valve function of these valves is discussed in paragraph 3.3.3. j

EE-22-0002 Rev. 1 p.10 3.3.1 (cont.)

The valves in the emergency condensate header are all remote manual. The firewater supply valve (HV 31122), the emergency condensate supply valves to the steam generator superheater sections (HV 2237, HV 2238), the emergency conden-sate supply valve to the steam generator reheater sections (HV 2291, HV 2290), as well as the emergency condensate flow control valves to the reheater section (FV 2239, FV 2240) are tested in accordance with specification SR 5.3.4, the frequency of which meets the required annual test frequency of criteria 3.2.2b of Ref. 1 for remote manual valves. The current test frequency is therefore adequate to demonstrate the operational readiness of these valves. It is recommended that testing of the emergency condensate header supply valves from the auxiliary boiler feedpump (HV 31121) and from the emergency condensate bypass (HV 31191) bo included in the surveillance procedure, since these valves should be closed in case of safe shutdown cooling.

(v) MISCELLANEOUS VALVES (TV2227, TV2228, HV2265, HV2266, HV22111 thru HV22114, HV22133, HV22134, HV22200 thru HV22212, HV22221 thru Hv22228, HV 21277)

There are miscellaneous safety related valves in the secondary coolant system. The main steam temperature control valves at the inlet of each EES module (TV 2227-1 through-6, TV 2228-1 through-6) are required to remain open for safe shutdown cooling. These valves are normally open, and fail safe in the open position. They are currently not tested, and criteria 3.2.2d of Ref. 1 exempt them from testing. Other safety related automatic or remote manual valves, such as the main steam steam trap isolation valves (HV 22221 through HV 22228), the cold reheat drain isolation valves (HV 22200 through HV 22212), the reheater attemperator block valves (HV 22133, HV 22134), the circulator nitrogen pressurization stop-check valves (HV 21277-4 through -7), the circulator cold reheat steam drain valves (HV 22111 through HV 22114),

and the reheater radiation monitor sample line block valves (HV 2265, HV 2266) are not currently tested. A failure of some of these valves to close, due to their small size (ranging from 3/4 inch to 2.5 inch) is not expected to be detrimental to the safe shutdown cooling capability of the system. The major safety function of these valves is then to provide a containment boundary (loop isolation) in case of a rupture in the steam generator. Accordingly, they are assigned to surveillance class S3, and the annual test requirement of criteria 3.2.2c for automatic valves is satisfied by normal operation since the valves are operated for plant startup and shutdown. Therefore, no surveillance is recommended for these valves.

l j (vi) The review of the valves under items (iii) through (v) above confirms that the requirements for surveillance testing

EE-22-0002 Rev. 1 p.ll 3.3.1 (cont.)

the safe shutdown cooling valves in the secondary coolant system allow technical specification SR 5.3.4 to be modified to provide additional operational flexibility, as indicated under item (ii). It is therefore recommended that SR 5.3.4 be modified to require that safe shutdown cooling valve testing be performed at the next scheduled reactor shutdown if the test was not performed during the previous year. This re-commended modification will be confirmed after reviewing the surveillance requirements for safe shutdown cooling valves in other systems.

b.3 Review of the rules for valve testing:

The existing rules for testing the above valves have been reviewed. Surveillance procedure SR 5.3.4-SA states that, in each step of the test, it is imperative that the operator be assured that the valve moves into the indicated position. This is performed by physically observing the valve position to con-firm the position indicating light. This method meets the proposed Code requirements, as outlined in paragraph (c) hereafter.

The current surveillance procedures do not provide for observation of the fail safe position of the valves, since almost all valves being tested are fail as is upon loss of actuator power. The following valves have been identified as having a fail safe position: the main steam depressurization valves, fail closed but are normally closed; the emergency condensate flow control valves to the reheater, fail closed but are normally closed; and the reheater discharge bypass control valves, fail open but are normally open. It is therefore considered, due to the nature and function of these valves,that no requirements for observation of valve fail safe position need be included in the technical specifications.

The surveillance procedures, except the one for the steam / water dump valves, do not provide for measuring the '

stroking time of the power operated valves. This is required by the test rules of the proposed Code and, therefore to meet the criteria of Ref. 1. However, this test requirement can be limited to the automatic hydraulic valves, since they are the only ones for which the stroking time may be of importance for performance of their safety function and since the high energy hydraulic actuators may be more susceptible to degra-dation than other types of actuators.

c) Proposed ASME Code surveillance requirements- l Seat leakage is not important for any of the valves under review. Therefore, only the proposed Code requirements appli-cable for Category B valves were considered. Paragraph

IGV-3300 requires that valves with remote position indicators

EE-22-0002 Rev. 1 p.12

3.3.1 (cont.)

be visually observed to confirm that remote valve indications accurately reflect valve operatior.. Paragraph IGV-3411 re-quires that the valves be exercised at least every 3 months.

However, paragraph IGV-3412 .llows valve exercising during 3

each cold shutdown, when a valve cannot be exercised during plant operation, and requires part-stroke exercising during plant operation if practical. Paragraph IGV-3413 requires that the plant owner specify the limiting value of full stroke time of each power operated valve, and that this time be measured whenever such a valve is full stroke tested. Paragraph IGV-3414 exempts from testing those valves which are exercised during plant operation at a frequency which otherwise satisfies the j exercising requirements of the Code. IGV-3415 requires that,

when practical, valves with fail-cafe actuators be tested by observing their operation upon loss of actuator power.

d) The recommended surveillance generally meets the proposed Code requirements with respect to frequency and test rules.

When the test frequency of the Ref. 1 criteria differs from

the test frequency required by the Code, such differences were justified by the recognition of the relative importance of the valves in performing the required safety funccions, and of the specific safety features of the Fort St. Vrain nuclear generating station.

i 1

3.3.2 SAFETY VALVES (V2214 thru V2216, V2245 thru V2247, V2225, V2262, V2270, V2275, V22170) i a) Current surveillance Iuirements:

There are no technical specifications related to sur-veillance testing of the secondary coolant systera safety valves. However, section 10.5 of the FSAR states that the safety valves throughout the secondary coolant system will be tested in accordance with applicable ASME Codes.

b) Recommended surveillance requirements:

! Criteria 3.2.2 (a, b and c) of Ref. 1 require that safety l valves be tested at a frequency and according to rules based on the proposed Code requirements, which is consistent with the FSAR. It is therefore recommended that a technical speci-fication be created to specify the surveillance requirements for the main steam, reheater and steam / water dump tank safety valves. It is recommended that testing of the safety valves be scheduled, when practical, so that valves of the same type operating under similar conditions are tested at different times during the surveillance interval, thus providing additional assurance with regard to the reliability of the overpressure protection throughout the. interval.

l

EE-22-0002 Rev. 1 p.13 3.3.2 (cont.)

c) Proposed ASME Code requirements:

Paragraph IGV-3511 requires that each safety valve be tested at least once every five years and provides the rules for scheduling safety valve testing within this time interval.

Paragraph IGV-3512 refers to ASME PTC 25.2-1966 for the rules for set point testing.

d) The recommended surveillance meets the proposed Code requirements.

3.3.3 CHECK VALVES (V2201, V2240, V2244, V2256, V2257, V2264, V2268, V2294, V22101, V22144, V22370, V22371; HV31118 thru HV31120, V31870; V4566; V8489) a) Current curveillance requirements: None.

b) Recommended surveillance requirements:

Check valves in the secondary coolant system are required to operate in case of safe shutdown cooling. A review was performed of the importance of individual check valve function and of the possiblilty of testing them to meet the criteria of Ref. 1.

The check valves on the emergency feedwater supply (V 2201, V 2240) and on the emergency condensate supply (V 2257, V 2256) to the steam generator superheater sections, are normally closed and are required to open for safe shut-down cooling. These valves can be tested by observing the feed-water flow (FI-2205,FI-2206) when the reactor is shutdown while supplying feedwater or condensate through the respective header.

The check valves on the feedwater line (V 2294, V 2244) provide a backup for the feedwater block valves; therefore, no testing is recommended.

Check valve V 22370 isolates the outlet from the inlet of the fire water booster pumps. It is required to close when a booster pump is used to supply the helium circu-lator water turbine drives. This valve can be tested at shut-down by operating one of the booster pumps and observing that it can develop its head at zero flow. Testing of check valve V 22371 is not practical. However, since an alternate flow path using V 22370 can be established, this valve can be exempt from testing.

Check valves V 2264, V 22101 on the feedwater supply

!. lines to the cold reheat steam desuperheaters are required to i .-

EE-22-0002 Rev. 1 p.14 3.3.3 (cont.)

close in case of safe shutdown cooling. However, as for the iso-lation valves (HV22133 and HV22134-see 3.3.1.b2(v) above), a leak of these check valves is not likely to affect safe shutdown cooling. Furthermore, the path can be isolated by closing the manual desuperheater isolation valves (V 2289 through V 2193, V 22119 through V 22125). Therefore, these valves can be exempt from testing.

Check valves V 2268, V 22144 on the emergency condensate supply to the steam generator reheater sections are normally closed anu required to open for safe shutdown cooling. These check valves can be tested, with the plant shutdown when it is possible to flood the reheaters with condensate, by observing the flow in the emergency condensate header.

As indicated in paragraph 3.3.1.b2 above, the emergency feedwater header stop-check valves (HV 31118, HV 31119, HV 31120) are reviewed as check valves since their controls are not safety related. These valves are required to close in case of safe shutdown cooling, but cannot be practically tested. However, since they have a manual override, they can be exempt from testing.

Check valve V 4566 on the firewater supply to the emer-gency condensate header is required to open in case of safe shutdown cooling. There is no practical way to test this valve without introducing firewater into the condensate system.

Check valve V S489 on the emergency condensate header supply from the auxiliary boiler feed pump is required to open for water supply, but since this supply is not safety related, its safety function is to close when the supply is from the firewater system. It is a backup for the isolation valve (HV 31121) end, therefore, can be exempt from testing.

The last safety related check valve (V 31870), on the emergency feedwater supply valve bypass from the motor driven boiler feed pump, is required to close. However, due to its small size and the fact that backup isolation is provided by normal valve V 31869, this check valve can be exempt from testing.

It is recommended that the normally closed check valves required to open to supply firewater for safe shutdown cooling j (through the emergency feedwater header or the emergency con-densate header to the steam generator superheater or reheater sections, and to the helium circulator water drives) be tested, if practical using condensate or feedwater, once a year with the plant shutdown or at the next scheduled shutdoms if such valves have not been tested during the previous year, to verify the operational readiness of the safe shutdown cooling system. l 1

t-

EE-22-0002 Rev. 1 p.15 3.3.3 (cont.)

c) Proposed ASME Code requirements:

• Paragraph IGV-3521 requires that check valves be generally exercised at least once every 3 months. Paragraph IGV-3522 allows check valve exercising to the position required to perform their safety function at each cold shutdown when exercising during operation is not practical. Normally closed valves (which is the case for the valves under consideration) required to open on reversal of pressure differential are to be tested either by proving that the flow is initiated or by applying a mechanical opening force to the disc.

d) The recommended surveillance meets the intent of the proposed Code for those check valves which can be tested. The reasons for exempting some valves from testing, as outlined in paragraph (b) above, justify the differences with the pro-posed Code requirements.

3.3.4 MANUAL VALVES a) Current surveillance requirements: .None.

b) Recommended surveillance requirements:

All the manual valves in the secondary coolant system are either instrument valves, vent and drain valves, maintenance or test valves. No surveillance is recommended for such valves.

c) Proposed ASME Code requirements:

Paragraph IGV-1200 exempts from testing the valves used only for operating convenience, such as manual vent, drain, instrument, and test valves, and valves used only for main-tenance.

d) The recommended surveillance and the proposed Code are consistent.

i 3.4 OPERATIONAL READINESS OF INSTRUMENTATION AND CONTROLS  !

l 1

A review of the instrumentation and control circuits in I system diagrams PI-22-1 through PI-22-9 has indicated that all the circuits, except for the ones discussed below, have either overall plant control functions, plant protective functions, or radiation monitoring functions. The surveillance for these circuits will be reviewed in separate reports, together with ,

system 93. j

,_ - . . - , - . . l

EE-22-0002 Rev. 1 p.16 3.4.1 STEAM WATER DUMP TANK INSTRUMENTATION (LT 2285, LT 2287, PT 2281, PT 22123, ii 2283, TT 22125, LSL 22155) a) Current surveillance requirements:

Technical specification SR 5.3.1 requires that the steam /

water dump tank level indicators be functionally tested monthly, and calibrated at each refueling.

b) Recommended surveillance requirements:

Technical specification LCO 4.3.3 does not allow plant operation at power if the steam / water dump tank condensate inventory exceeds a predetermined limit (about 2100 gallons, corresponding to a level of 45 inches). This prevents lifting the tank safety valves, should a loop dump occur, and releasing some radioactivity to the environment due to the reactor coolant mixed with the steam and water being dumped and leaking through the ruptured steam generator tube or subheader. The steam water dump tank is equipped with redundant level measurements (LT/LI/LS/LIHL-2285 and -2287) to monitor the water level. The current surveillance exceeds criteria 3.2.3 of Ref. 1 which requires annual functional testing and calibration for sur-veillance class S4 instrumentation. However, since these instruments are used to verify a limiting condition for opera-tion, it is recommended that the functional test frequency be reduced from monthly to quarterly (rather than to annually),

to limit the amount of 11guid waste generated by the functional test. This test is performed by increasing the water level, then restoring it to normal, with the condensate used for the test being sent to the radioactive liquid waste system. The

! instrument calibration frequency may not meet the criteria of Ref. 1 when the refueling cycle exceeds one year. However, since the dump tank is not easily accessible during plant operation and due to the risk of a water / dump occurring while calibrating the level transmitter, such calibration should be performed with the reactor shutdown. It is therefore recommended that the technical specification be modified to require dump tank level instruments calibration annually, or at the next scheduled plant shutdown, if it has not been performed during the previous year.

Redundant pressure instruments (PT/PI/PSH/PIH-2281 and

-22123), in combination with the dump tank radiation monitors, provide the operator with information confirming that a leaking steam generator has been dumped, since the pressure and radiation will increase, due to inleakage of reactor coolant, above the values expected in case of wrong loop dump. The pressure instruments also monitor that the tank gas phase is vented to the radioactive gaseous waste system. Redundant temperature instruments (TE/TT/TI/TSH/TIH-2283 and -22125) monitor the cooling of the tank inventory following a steam /

water dump, so that the tank level can be brought back to normal,~after draining and venting following adequate cooling.

EE-22-0002 Rev. 1 p.17 3.4.1 (cont.)

Such cooling.is required to prevent steam from being vented to the gas waste system, and hot water from being drained to the liquid waste system. An interlock from PSH-2281 prevents the dump tank drain pump (P 2201) from operating when the tank pressure exceeds 50 psig. Another interlock (LSL-22155) stops the pump in case of low water level in theSince tankthe to prevent above damage to the pump due to lack of NPSH.

instrumentation ir used to prevent damage to the radioactive waste systems, it is recommended that it be calibrated when calibrating the dump tank level instruments. The radiation monitors are not included in these requirements, as they will be reviewed in a separate report for system 93.

Proposed ASME Code requirements: Not applicable.

c) 3.4.2 VALVE INSTRUMENTATION AND CONTROLS Current surveillance requirements: None.

a) b) Recommended surveillance requirements:

The valve position indication circuits are functionally and-j

' tested when the valves are tested, and so are most of t.

switch controls. Some automatic controls are also teste l

when performing the valve test, which are not part of the plant protective system or of the overall plant control system.

These controls are identified below.

An interlock prevents opening the emergency condensate supply valve to a steam generator superheater section, as long as either the main feedwater block valve or the emergency feed-water supply valve is open; this interlock is tested per SR 5.3.4.

The pressure controllers for the main steam power operated pressure relief valves (PC-22167, PC-22168) are functionally tested when testing the valves per technical specification SR 5.3.3.

The emergency condensate flow controllers (FC-2239, FC-2240) to the reheaters are functionally tested when testing the flow control valves per technical specification SR 5.3.4.

However, the flow transmitters, modifiers and recorders are not tested or calibrated. These instruments and controls are re-quired to be functional in case of safe shutdown cooling with the reheater. They also provide the operator with an indication l

of the emergency condensate flow, which he can use for remote manual control of the safe shutdown cooling. It is therefore recommended that these instruments and controls be functionally tested and calibrated about once a year.

- . - , _ - . - - - mr--w -r - - -w~-- w =v

EE-22-0002 Rev. 1 p.18 3.4.2 (cont.)

The same considerations apply to the reheater discharge bypass pressure control circuits (PT/PC-2267 and -2268). The valve testing per SR 5.3.4 provides a functional test of the bypass block valve / bypass control valve interlock, and of the pressure controller. No functional test is currently performed for the pressure transmitter. It is recommended that the transmitter be also functionally tested and calibrated since it is used to prevent boiling in the reheater in case of safe shutdown cooling. It is also recommended that pressure switches PSH-22197 and PSH-22198 be functionally tested and calibrated since they provide overpressure protection for the reheaters, while preventing lifting of the safety valves.

c) Proposed ASME Code requirements: Not applicable.

3.4.3 REHEAT STEAM INSTRUMENTATION a) Current surveillance requirements: None.

b) Recommended surveillance requirements:

The following secondary coolant system instrumentation provides information which can be used by the plant operator to prevent damage to the reheaters or the circulators.

Pressure differential measurements across the strainers at the reheater inlet (PDT/PDI/ PDM /PDSH-22231-1 through -6, PDAH-22231; PDT/PDI/ PDM /PDSH-22232-1 through -6, PDAH-22232) provide the operator with information about strainer fouling and possible carryover into the reheaters. The same con-siderations apply for fouling of the circulator inlet pro-tective screens (PDT/PDI/ PDM /PS-22233 through -22236, PDAH-22233 and -22234). Temperature instruments (TE-2232-1 through -12, TE-2255-7 and TE-2256-7, TR-2232, TAL-2232) warn the operator about excessive desuperheating and possible carryover of moisture in the reheaters. Temperature instruments (TS/TAH-22195, -22194) warn the operator of excessive reheater dis-charge bypass temperature to protect the condenser from verheating.

The above instrumentation is assigned to surveillance class S4 for which annual functional testing and calibration are recommended.

c) Proposed ASME Code requirements: Not applicable.

-)

i

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EE-22-0002 Rev. 1 p.19

4. STRUCTURAL INTEGRITY 4.1 STEAM GENERATOR TUDES a) Current surveillance requirements:

There are no current technical specification surveillance requirements for the steam generator tubes. However, existing plant instrumentation continuously monitors the steam genera-tor heat transfer sections for evidence of tube degradation.

Moisture monitors installed in the reactor instrumentation and analytical instrumentation system will indicate, alarm and initiate necessary protective actions if a leak should develop in the superheater section. Radiation monitors instal'ed in _

the secondary coolant system hot reheat piping will indicate, alarm and initiate protective actions if a leak should develop in the reheater section. In addition, instrumentation in the analytical instrumentation system and the steam and water sample system continuously monitors primary and secondary coolant purity.

Existing plant instrumentation also continuously monitors the superheater sub-headers which are located within the PCRV penetration. Penetration pressure and moisture monitors installed in the PCRV auxiliary system indicate, alarm and initiate protective actions if a leak should develop in one of these sub-header sections. Reheater steam piping headers located within the PCRV penetration are continuously monitored as well. Penetrations purge flow monitors in the PCRV auxiliary system alarm and initiate protective actions if a leak should develop in these reheat piping sections, which are at a lower pressure than the penetration interspace.

b) Recommended surveillance requirements:

The once through type steam generators at Fort St. Vrain operate with reactor coolant (high purity helium) outside the tubes and secondary coolant (high purity steam / water) inside the tubes. Thick wall tubes, arranged in helical tube bundles, provide the required strength and flexibility to withstand the high operating pressures and temperatures.

Structural integrity of the steam generator tubes is required for safe shutdown cooling, where either one suocrheater

, or one reheater section and one of the corresponding circulators must remain operable or be restored to operation within the time delay allowed before a permanent loss of forced circula- ,

tion accident is declared. Structural integrity of the tubes j is also required since they comprise a portion of the reactor I coolant pressure boundary and therefore function as primary containment, one of the major barriers against release of

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

,,.,.,,,,,,,.y_._.7 _.-__,--wr ,<--,--m---

EE-22-0002 Rev. 1 p.20 4.1 (cont.)

fission products to the environment. The tubes also act as a barrier to prevent steam or water from entering the reactor coolant system and possibly damaging reactor internal components.

There currently is no method available for inspecting steam generator tubes without removing steam generator modules from the PCRV. The tubes are not accessible from the primary side due to the shroud design which surrounds the tubes and cannot be inspected internally, using current technology, because of the tube design (helical tube bundles, varying tube I.D. and 90 turns at the tube to header or subheader junctions). Although the PCRV was designed with provisions for removal and replacement of steam generator modules, it would be a difficult, costly and time consuming task. Further-more, the method has not been demonstrated nor is equipment available to do the job. Therefore, non-destructive examina-tion of the steam generator tubes is considered impractical i

at this time and cannot be used to verify tube integrity.

There are no provisions in the design of the PCRV for insertion of tube material specimens in the reactor which would provide meaningful and representative data regarding tube integrity since the primary concern is cyclic thermal stresses across the tube wall. Steam generator tube integrity must therefore be monitored indirectly.

Corrosion protection for the steam generators is dis-cussed in section 4.2.4.3.8 of the FSAR. Therein, it is shown that expected metal loss for the tube materials operating at specified conditions is insignificant.

Corrosion in boiler tubes, in general, is related to feedwater quality. Out-of-specification feedwater chemistry could lead to excessive deposition of contamination on the inside tube wall, which could accelerate corrosion. The results of a special program to monitor steam generator per-formance with respect to corrosion were reported in reference

18. Data monitored included tube side pressure drop, heat transfer section temperatures and feedwater chemistry. Data i

analysis showed that steam generator performance was as ex-pected; no evidence of deposit buildup was observed. Reference 18 also reported the results of examinations performed on.two feedwater ring header trim valves, one feedwater ring head'er strainer and portions of the feedwater and steam leads which were removed to plua a leaking tube. Again, no unusual corrosion conditions were observed. These findings tend to verify the design expectations that corrosion and subsequent tube thinning is of little concern for the Fort St. Vrain steam generators, provided that primary and secondary coolant purity is main-tained within specified limits. Since the primary and secondary coolant chemistry is continuously monitored there is no need

, for additional corrosion surveillance and none is recommended.

EE-22-0002 Rev. 1 p.21 4.1 (cont.)

Vibration and wear protection of the steam generator tubing is described in sections 4.2.4.3.4 and 4.2.4.3.5 of the PSAR. The desiq, provisions for wear protection are based on proven technology and do not present any concerns with respect to tube thinning. The design provisions for vibration have been confirmed by the plant startup test pro-gram wherein one steam generator module was fully instrumented with strain gages. These measurements have not indicated any cause for concern relative to tube vibration. Therefore no surveillance is considered necessary to further monitor for tube motion.

Design of the tubes to withstand the stresses imposed by operating conditions is also described in section 4.2.4 of the FSAR as well as in the steam generator design report.

The analysis met all applicable Code requirements in existance at that time. Since then, several potential concerns related to tube life have been identified, based on more recent knowledge of creep-fatigue interaction and cold working of Incoloy 800 tubing. Additional research and development pro-grams are necessary to determine if these concerns could actually affect predicted tube life. However, since the tubes cannot be directly inspected, defects can only be identified after they have propogated to the point where a leak is initiated.

A review of the current design provisions, as described in (a)-

above for monitoring tube leakage, indicates that there is sufficient instrumentation provided to detect both small and large leaks. Technical Specifications LCO 4.2.10/4.2.11 also specify limits for primary coolant impurities, which would include moisture entering the reactor via a leaking superheater tube. Technical Specification LCO 4.3.8 specifies limits for secondary coolant activity, which would likely be indicative of a reheater tube leak. These provisions are considered adequate to verify the integrity of the steam generator tubes and meet criteria 3.3.2 (c) of reference 1 for continuous leakage monitoring. Surveillance requirements for the instru-mentation and control circuits which are used to verify tube integrity will be addressed in separate reports covering the plant protective and radiation monitoring systems (System 93) .

There has been one superheater tube failure to date. It was identified by the instrumentation described above and was subsequently located and successfully plugged. The vertical location of the leak, as determined by indirect means, was not located where failures would be expected due to the potential concerns previously mentioned and was therefore considered a random failure. Experience to date therefore does not justify that additional surveillance be considered.

Regulatory Guide 1.83 has been reviewed as a reference document, to determine if some requirements could apply to the Fort St. Vrain steam generators, even though the Guide specifically addresses inservice inspection of PWR steam generator tubes. It was determined that the differences in design are cuch that Regulatory Guide 1.83 cannot be considered

1 l

EE-22-0002 Rev. 1 l p.22 4.1 (cont.)

1 c) Proposed ASME Code requirements:

The inspection and test requirements of subsection IGB would apply to the steam generator, which is considered to be Code Class 1 since it forms part of the reactor coolant pres-sure boundary and functions as primary containment.

Paragraph IGB-2510 requires components to be examined for leakage following each reactor refueling outage and to be pressure tested and examined for leakage at or near the end of each inspection interval (approximately every ten years). The required test pressure for the periodic leakaoe examination is nominal system operating pressure at 100%

rated reactor power. The required test pressure for the system pressure test is 1.25 times the system design pressure.

Paragraph IGB-2520 and IGB-2600 require that components be non-destructively examined for defects as specified in Tables IGB-2500-1 and IGB-2600-1. There are no requirements specified in these tables for steam generator tubes.

Subsection IGX, which is not yet available, will provide requirements for components subject to elevated temperature service, which would be the case for the steam generator tubes. It is understood that these requirements will generally provide for surveillance using material specimens.

d) The current and recommended surveillance requirements exceed the proposed Code requirements for leakage examination, since leakage is continuously monitored at power. For this same reason and since the tubes are not accessible for leakage examination, pressure testing would not provide additional assurance of integrity and may not be practical for this design.

4.2 STEAM GENERATOR CLOSURES The structural integrity of those components of the steam generators which functions as primary and secondary closures for the PCRV penetrations has been reviewed in the reports on the PCRV and the PCRV Auxiliary Systems (references EE-ll-0001 and EE-ll-0002).

4.3 STEAM /WA*iMR DUMP TANK STRUCTURAL INTEGRITY I

a) Current surveillance requirements: None.

l l

I

EE-22-0002 Rev. 1 p.23 4.3 (cont.)

b) Recommended surveillance requirements:

Structural integrity of the dump tank is required during the short period when a steam / water dump is performed, to contain the primary reactor coolant which leaks through a failed steam generator tube or subheader. After the dump is completed, the dump valves are closed and act as containment boundary. During normal operation of the plant, the dump tank is not subjected to any conditions expected to dearade its integrity when compared to design conditions. Further, the dump tank instrumentation continuously monitors the leak tightness of the tank (level monitoring for water leaks, pressure monitoring for gas leaks). Therefore, no additional surveillance is recommended to verify the structural integrity of the steam / water dump tank.

c) Proposed ASME Code requirements:

The steam / water dump tank is considered to be Code class 3, since it may contain radioactive fluido, and the requirements of subsection IGD are deemed applicable. Para-graph IGD-5210 requires that a pressure test and an examination for leakage be performed at the end of each inspection interval (about every 10 years) . Paragraph IGD-5220 requires that the test pressure be at least 1.10 times the system design pressure.

d) Differences exist between the recommended surveillance and the oroposed Code requirements,which are considered justified for the reasons outlined in paragraph (b) above.

4.4 STRUCTURAL INTEGRITY OF SYSTEM PIPING a) Current surveillance requirements: None.

b) Recommended surveillance requirements:

b.1) Leakage monitoring:

Structural integrity of the safety class I secondary coolant piping is required to perform the safe shutdown cooling mode of operation for which either a superheater or a' reheater, and one of the corresponding circulators, need to remain operational, or be restored to operation within the time delay allowed before an accident is considered as a per-manent loss of forced circulation accident. Structural integrity j of the secondary coolant piping is also required to provide a j secondary containment, in case of a steam generator leak in 1 the primary reactor coolant system, between the time such a  !

leak is detected and the time the reactor coolant system is j depressurized to subatmospheric pressure.

1 i

i

EE-22-0002 P. 24 s

4.4 (cont.)

The plant is provided with a safety related steam pipe rupture detection system (system 93-5) which continuously monitors secondary coolant piping structural integrity within the reactor building. This system alerts the operator, in case of small leaks, or automatically trips one loop and scrams the reactor, in case of significant leaks for which operator action would be considered too slow. Surveillance of the steam pipe rupture detection system will be addressed in a separate report.

Other instrumentation also provides for additional continuous leakage monitoring of the secondary coolant system throughout the plant. Such instrumentation monitors the feedwater and emergency feedwater flow, the main steam and hot reheat steam pressure, and the circulator speed. This instrumentation is adequate to monitor large pipe leaks. Smaller leaks will be detected and monitored by plant operators when performing daily routine plant inspections, and corrective actions will be taken before the damage becomes critical.

Therefore, it is considered that existing leakage monitoring meets criteria 3.3.la of Ref.1, and is adequate to demonstrate the structural integrity of the secondary coolant system piping.

B.2) Examination:

Due to the importance of the secondary coolant system piping in the several safety functions, a review was performed of the design reports to determine if the stress levels, under all the postulated normal and accident conditions, would raise a concern that a common failure might occur which in turn could justify additional examination. The thermal stresses and the safe shutdown earthquake stresses were found well within allowable limits. In a few instances, the fatigue stresses were found somewhat high, when determined by elastic analysis, but acceptable when more refined analysis methods were used. Another potential area of concern could have been corrosion induced pipe thinning which might degrade pipe strength. However, as already outlined in paragraph 4.1 above for the steam generators, no such phenomenon has been observed at Fort St. Vrain due to the controls required for feedwater chemistry and the materials used in the system.

Therefore, no concerns were identified from a stress or corrosion point of view which might warrant that examination of the secondary coolant piping be recommended for the surveillance program of the Fort St. Vrain nuclear generating station.

l

W EE-22-0002 Rev. 1 l p.25 4.4 (cont.) l A review was performed of the potential consequences of cracks occuring in the secondary coolant system piping, for reasons similar to or different from the ones which caused cracks in piping of other nuclear power plants. Cracks of unknown origin will be first identified when pipe leakage occurs, when no other in-service examination is performed.

It is not expected, based on experience in nuclear and con-ventional power plants, that the phenomenon causing such cracks, would result in common mode shear pipe rupture, even in the event of a safe shutdown earthquake, which might pre-vent safe shutdown cooling, without having shovn early signs of degradation resulting in leakage. Even if Ton-destructive examinations had been performed, such phenomenon could occur during the inspection interval and may not have been identified during the examination, in particular if it were fatigue induced cracking. Some concern could be raised if such early signs of leakage could remain undetected over a period of time. This could be the case in other nuclear power plants, where the safety related piping is located within a containment building where access is not allowed during plant operation at power, so that a small leak could remain undetected for as long as a refueling cycle. However, the entire secondary coolant system of the Fort St. Vrain nuclear generating station is accessible at all times and the plant operators perform daily routine inspections so that such early signs of leakage will be readily detected. Waiting for early signs of leakage is consi-dered acceptable at Fort St. Vrain since the design provides _

diversity of cooling methods, in addition to redundancy, so that, even if a problem appears in one kind of oipe, it is -

unlikely that the same problem would occur in another kind of pipe with different operating conditions and layout, and thus safe shutdown cooling will always remain possible.

It is therefore recommended that additional examinations be considered only when leaks develop which may raise a concern with respect to structural integrity of the secondary coolant piping. It should be noted that no leakage of the secondary coolant system pressure boundary has been experienced thus far.

c) Proposed ASME Code requirements:

The inspection and test requirements of subsection IGC would apply to the safety related secondary coolant system piping, which is considered to be Code Class 2 since it is used for safe shutdown cooling and as a containment boundary.

Paragraph IGC-2510 requires that fluid system comnonents be pressure tested and examined for leakage at or near the end of each inspection interval (approximately every ten years) .

The required test pressure is 1.25 times the system design pressure.

EE-22-0002 Rev. 1 p.26 4.4 (cont.)

Paragraph IGC-2520 and IGC-2600 require that fluid system components be non-destructively examined for defects as spe-cified in Tables IGC-2500-1 and IGC-2600-1.

Table IGC-2500-1, category C-G,and Table IGC-2600-1 re-quires that 50% of the total number of certain pressure retaining weld joints in piping system components be subject to volumetric examination. Table IGC-2500-1,. category C-D,and Table IGC-2600-1 require that all pressure retaining bolting greater than one inch in diameter be subject to visual examination and that 10% of the bolting in each joint be subject to either surface or volumetric examination. These examinations are required to be performed on a scheduled periodic basis during each inspection interval (approximately every ten years).

However, examination of bushings, threads and ligaments in the base material of flanges are required only when a bolto' connection is disassembled'for other reasons.

d) The differences between the recommended surveillance and the proposed Code requirements are justified by the considera-tions of paragraph (b) above which are based on the particular safety features of the Fort St. Vrain nuclear generating station.

5. LIST OF REFERENCES

References:

(1) PSC report EE-SR-0001: Surveillance inspection and test criteria for the Fort St. Vrain nuclear generating station.

(2) Fort St. Vrain FSAR: Sections 4, 6.2, 6.3, 6.4, 10.

~

(3) FSV technical specifications LCO 4.2.10, LCO 4.2.11, LCO 4.3.1 through LCO 4.3.4, LCO 4.3.8, SR 5.3.1 through SR 5.3.4, SR 5.3.7.

(4) FSV surveillance procedures SR 5.3.1-M, SR 5.3.1-Q, SR 5.3.1-R, SR 5.3.2-W, SR 5.3.2-A/5.3.3b-A, SR 5.3.3a-A, SR 5.3.4-SA/5.3.3c-a, SR 5. 3. 7-W.

(5) _FSV system diagrams: PI-22-1 through -10, PI-31-1, SR G-1 sheets 28 through 35 and sheet 60.

(6) _ FSV system electrical diagrams: E-1203 pages 164, 390, 391, 401 through 481, 950, 951, 1221 through 1223, 1620 through 1629, 1860 through 1952, 1986, 1987, 1923A, 1986A, 1987A.

(7) FSV drawings: B2201-100, 101, -240, -630, -750, -790; C2101-422, -425, -426, -430, -431, -432; M-35 through -65, M-325 through -334, M-343 through -357, M-9102-150, -160,

-170, -180; R-1100-100.

9 EE-22-0002 Rev. 1 p.27

5. (cont.)

(8) FSV design criteria and system descriptions: DC-22, SD-22-1, SD-93-5.

(9) PSC letter P-78071, J.K. Fuller to R.P. Denise, dated May 8, 1978: Manual Class I/ Class II Valve Study.

(10) FSV equipment specifications: 90-SR-6-2, 91-M-01, 91-M-13, 91-M-14, 91-M-16, 91-M-19, 91-M-23, 91-M-51A, 91-M-57, 93-I-5, 91-M-5.

(11) FSV equipment drawings: 90/91-M-13-40, -52, -53, -70,

-105, -152; 91-M-5-27 through -36; 1-M-2-1707, 91-M-19-1 through -16, -35, -43, -44, -65 through -68.

(12) Draft ASME Code,Section XI, Division 2.

(13) Attachment C to Amendment 26 of the plant operating license.

(14) Stress analysis reports SL-2783, SL-2841, GADR-134, GADR-135.

(15) Control logic diagrams IB-93-6, -7, -8.

(16) Control and instrument diagram IC-93-4.

(17) NRC Regulatory Guide 1.83, Rev. 1, dated July 1975.

(18) PSC letter P-79058, J.K. Fuller to W.P. Gammill dated March 15, 1979: In-Service Inspection Program for Fort St. Vrain.

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