ML20205M059

Forwards Summary of Audit of Changes Made to Facility Under 10CFR50.59.Concerns Noted Include Safety Basis in Some Change Notices Not as Complete as Desirable & Accurate Safety Evaluations Could Not Be Developed from Documents

ML20205M059
Person / Time
Site:	Fort Saint Vrain
Issue date:	03/25/1987
From:	Heitner K Office of Nuclear Reactor Regulation
To:	Robert Williams PUBLIC SERVICE CO. OF COLORADO
References
TAC-62126, NUDOCS 8704020256
Download: ML20205M059 (17)

Text

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y l 7, 'j WASHINGTON, D. C. 20555. l MAR 2 51987 Docket No. 50-267 Mr. R. O. Williams, Jr.

Vice President, Nuclear Operations Public Service Company of Colorado P. O. Box 840 Denver, Colorado 80201-0840

SUBJECT:

AUDIT OF CHANGES TO FORT ST. VRAIN MADE UNDER 10 CFR 50.59 We have completed an audit of changes made by Public Service Company of Colorado (PSC) to the Fort St. Vrain Nuclear Generating Station (FSV) under the provisions of 10 CFR 50.59. PSC reported these changes to the Nuclear Regulatory Commission (NRC) in a letter dated July 22, 1985 (P-86454).

This audit was conducted by the Office of Nuclear Reactor Regulation (NRR) because of a general concern that 10 CFR 50.59 reviews were not always correctly conducted by licensees. Although a licensee's activities in this area are inspected by other offices of the NRC, the Director of NRR determined that this activity should also be audited by the NRR staff. Specifically, he designated this responsibility to each plant's Project Manager (s). In order to fulfill this responsibility, on December 16, 1986, Mr. Charles S. Hinson and I, Project Managers for FSV, visited the PSC offices at Diamond Hill, where we performed an audit of nine selected change notices (CNs). A listing of the CNs audited and summaries of the results of this audit are contained in Enclosure 1.

Our goal in this audit was to provide a broad overview of the process PSC used to make changes under 10 CFR 50.59. Since NRR has not formalized this audit process in the manner of plant inspections, we are providing the results of this audit to you for information only. No formal response to the audit observations is required.

Our audit resulted in three major concerns. First, the safety bases in some of the CNs were not as thorough or complete as might be desirable.

Specifically, a clear understanding of the basis for finding the change acceptable under the provisions of 10 CFR 50.59 could not be obtained by reading the CNs. Potentially, inadequate quality standards for CNs could lead to erroneous determinations. We recomend that PSC review the CN process and be assured that acceptable quality is maintained in this activity.

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MAR 2 51987 i Mr. R. O. Williams, Jr. '

Second, we observed that accurate safety evaluations for the CNs could'not-o always be developed from-the existing documentation that defines-FSV's

( licensing basis. Specifically, we observed in CN-1272 A and B, that a complete ,

i evaluation of the capacity for the safety related station batteries was not '

available. These deficiencies include both the existing Technical Specifications

,' and the updated FSAR. We, of course, acknowledge PSC's efforts to improve the

quality of these~ documents, especially through the Technical Specifications

Upgrade Program (TSUP). We recommend PSC continue these' efforts through the 1

FSAR update process,"and by consulting the improved TSUP bases.

, Third, we observed that the PSC staff did not always evaluate the full set of possible technical consequences from the proposed modification. For example, in CN-2038 the.PSC staff only considered the consequences ~of alternate modes of the bearing. water pump's normal operation in evaluating the impact of the change. However, consequences of abnormal operation from failure of all three i pumps were not evaluated in the CN. The PSC staff noted that other valves in the system would prevent back flow if all three pumps failed. Although, we agreed with PSC's final conclusion, we again observed that this was because i other components served to perform the necessary protective function, and.

l these were not discussed in the CN.

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We discussed our observations with the PSC staff members listed in Enclosure 2.

Their comments helped us to understand better some of the CNs. They also pointed out that several of the CNs were processed up to two years ago. PSC believed that improvements made under the Performance Enhancement Program responded to some of our recommendations.

} If you have any questions about the results of this audit, please call me at 1 (301)492-8205.

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Sincerely, Ortdinni Signed By 0 i Kenneth L. Heitner, Project Manager Standardization and Special i Projects Directorate a Division of PWR Licensing-B

~ Office of Nuclear Reactor Regulation

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Mr. R. O. Williams, Jr. Second, we observed that accurate safety evaluations for the CNs could not always be developed from the existing documentation that defines FSV's '

licersing basis. Specifically, we observed in CN-1272 A and B, that a complete evaluation of the capacity for the safety related station batteries was not available. These deficiencies include both the existing: Technical Specifications and the updated FSAR. We, of course, acknowledge PSC's efforts to improve the quality of these documents, especially through the. Technical Specifications

. Upgrade Frogram (TSUP). We recommend PSC continue these efforts through the FSAR update process, and by consulting the improved TSUP bases.

Third, we observed that the PSC staff did not always evaluate' the full set of possible technical consequences from the proposed modificatien. For exarrple, in CN-2038 the PSC staff only considered the consequences of alternate modess of the bearing water pump's normal operation in evaluating the impact of the h change. However, consequences of abnormal operation from failure of all. three ,

pumps were not evaluated in the CN. The PSC staff noted that other valves in 4 the system would prevent back flow if all three pumps failed. Although, we '"

agreed with PSC's final conclusion, we again observed that this was because -

4 other components served to perform the necessary protective function, and these were not discussed in the CN.

We discussed our observations with the PSC staff members listed in Enclosure 2.

Their coments helped us to understand better some of the CNs. They.also. , ,

pointed out that several of the CNs were processed up to two years ago.. PSC believed that improvements made under the Performance Enhancement. Program responded to some of our recommendations. -

l If you have any questions about the results of this audit, please call me at (301)492-8205. .

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Sincerely,

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Kenneth L. Heitner, Project Manager Standardization and Special Projects Directorate Division of PWR Licensing-B Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next page

Mr. R. O. Williams Public Service Company of Colorado Fort St. Vrain cc:

Mr. D. W. Warembourg, Manager Albert J. Hazle, Director Nuclear Engineering Division Radiation Control Division i Public Service Company Department of Health of Colorado 4210 East lith Avenue P. O. Box 840 Denver, Colorado 80220 Denver, Colorado 80201 Mr. David Alberstein, 14/159A Mr. R. O. Williams, Acting Manager GA Technologies, Inc. Nuclear Production Division Post Office Box 85608 Public Service Company of Colorado San Diego, California 02138 16805 Weld County Road 19-1/2 Platteville, Colorado 80651 Mr. H. L. Brey, Manager Nuclear Licensing and Fuel Division Mr. P. F. Tomlinson, Manager Public Service Company of Colorado Quality Assurance Division P. O. Box 840 Public Service Company of Colorado Denver, Colorado 80201 16805 Weld County Read 19-1/2 Platteville, Colorado 80651 Senior Resident Inspector U.S. Nuclear Regulatory Commission Mr. R. F. Walker P. 0. Box 840 Public Service Company of Colorado Platteville, Colorado 80651 Post Office Box 840 Denver, Colorade 80201-0840 Kelley, Stansfield & 0'Donnell

"; Public Service Company Building Comitment Control Program Room 900 Coordinator 550 15th Street Public Service Company of Colorado Denver, Colorado 80202 2420 W. 26th Ave. Suite 100-D Denver, Colorado 80211 Regional Administrator, Region IV U.S. Nuclear Regulatory Commission 611 Ryan Plaza Drive Suite 1000 Arlington, Texas 76011 Chainnan, Boarti of County Commissioners of Weld County, Colorado Greeley, Colorado 80631 Regional Representative Radiation Programs Environmental Protection Agency 1 Denver Place 999 18th Street, Suite 1300 Denver, Colorado 80202-2413

I Enclosure 1 SlfMMARIES OF FORT ST. VRAIN CHANGE NOTICES AUDITED ON DECEMBER 16, 1986

1. CN-1624 - System 11 - Reactor Vessel and Internal Components.

2. CN-1876 - System 12 - Control Rod Drive and Orfficing Assembly.

3. CN-1983 - Systems 21 & 22 - Primary Coolant System and Secondary Cooling System.

4 CN-1889 - System 11 - Reactor Vessel and Internal Components.

5. CN-1272 A, B - System 92 - Electric Power System.

6. CN-1977 - System 92 - Electric Power System.

7. CN-1979 - System 92 - Electric Power System.

8. CN-1996 - System 92 - Electric Power System.

9. CN-2038 - System 21 - Primary Coolant System.

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1. CN-1624 - System 11 - Reactor Vessel and Internal Components 4_

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1. Why Selected The Fort St. Vrain' moisture monitors are a critical system to insure

proper plant operation. However, they have not' performed well. This

modification to the monitars appeared significant and was therefore chosen for review.

! 2. Summary of Change

! This change notice (CN) involvec replacement of moisture monitor isolation 1

valves in Instrument Penetrations B-1 through B-6 and the use of

}' compression type fittings in one-inch 0.D. and smaller process and' instrument tubing. The moisture monitor isolation valves were replaced with valves having a secondary sealing system to limit the leakage of helium into the penetration's secondary interspaces in the event of 4 failures in the valve's bellows seals and valve seats. Other fittings

! and tubing changes were made to increase the accessibility of the-moisture monitors for calibration, maintenance and removal.

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3. Tests for Significance This change was safety related and resulted in changes to two FSAR

figures. However, it was not determined to be safety significant in that -

i these changes will not affect the capability to prevent or mitigate .the

! consequences of accidents described in the FSAR.

I The rerouting of the moisture monitoring tubing was determined not to be

, an unreviewed safety question because: 1) this change will not increase the response time of the high level moisture monitors and hence will not

affect any equipment important to safety, 2) the function of the high level moisture monitors has not been changed, and 3) the response time of

, the high level moisture monitors has not been increased and therefore no j margin of safety has been affected.

i This change involves systems and components described in Technical

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Specifications (TS) Sections LCO 4.4.1 (Instrumentation and Control  ;

Systems) and LSSS 3.3. However these changes do not involve any changes l to the above TS. These changes also do'not require any change in the >

. frequency of the surveillance and calibration requirements for .the j primary coolant moisture monitor channels.

i 4. NRC Evaluation and Conclusions l The staff agrees with the licensee's evaluation of these changes. The licensee performed a comprehensive evaluation of these changes including

safety and mechanical analyses. On the basis of our review, we determine

that these changes were within the secpe of 10 CFR 50.59 and do not ,

! result in any margin of safety being affected.

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2. CN-1876 - System 12 - Control Rod Drive and Orificing Assembly

1. Why Selected The change affects one of the basic reactor parameters, coolant flow. It represents a change to the orificing valve system, which in turn controls' flow of coolant through the reactor.

2. Sumary of the Change This modification allows the continuous movement of the orifice valve over its allowable range of motion (6 inches). Previously, an interlock prevented continuous movement of the orifice valve by requiring the- hand switch to be reactivated every 0.0125 inches of orifice valve travel.

Since the interlock and time delay permitted only one increment of valve travel per 5 seconds, the total time to move the valve the 6 inches of travel was 41.67 minutes. With removal of the interlocks, the valve can be moved the 6 inches of travel in only 30 minutes. TS require that orifice valves be set for equal region flows at lower temperatures and that they be set for equal region outlet temperatures at higher temperatures.

This modification will reduce the time required to reorifice the core, in order to meet TS limits, during reactor power level changes and reactor runback transients.

An additional change that was made was the addition of limit trips to prevent the valve from running into the mechanical stops at each end of travel. These limit trips will reduce the possibility of an orifice valve becoming stuck at the stops.

3. Tests for Significance This change is safety related in that the systems involved are Class 1 and affect safe shutdown. This change was also determined to be safety significant in that the modification involves changes to safety related systems.

The removal, in part, of interlocks and time delays so that continuous travel of the orifice valves can be accomplished was determined by the licensee not to be an unreviewed safety question on the basis of the following analyses. Since deliberate operator action is required in both the existing design and the design change to fully open or fully shut an orifice valve, the probability of occurrence and the consequences of this accident, which was previously evaluated in the FSAR, will not be increased. Increasing the speed at which the orifice valve travels could increase the possibility of sticking a valve at the mechanical stops if the valve were to be driven into one of its mechanical stops. The FSAR does not evaluate an orifice valve stuck in any position. Installation of trip limits at the open and shut positions decreases the probability 1

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l Page 4 of an orifice valve being driven into either mechanical stop. To provide adequate insurance that a stuck valve will not occur, administrative controls shall be in place to require the operator to stop valve travel before reaching the trip limits. Provided that the administrative controls on valve repositioning and periodic tests on the automatic trip limits are in place, this design change does not create the malfunction of a sticking orifice valve v.hich would result if the valve were driven into its mechanical stop. There is no margin of safety or orifice valve repositioning time. Although the increased orifice valve speed may increase the frequency wherewith operators must take corrective actions to comply with LC0 4.1.7 and LC0 4.1.9, compliance with these LCOs ensures that fuel damage will not occur.

This modification requires no changes to the applicable sections of the TS (LCO 4.1.7 (Core Inlet Orifice Valves) and LCO 4.1.9 (Core Region Temperature Rise)). This modification will allow the operators to be better equipped to maintain core outlet temperatures within the requirements of LC0 4.1.7 and 4.1.9.

4 NRC Evaluation and Conclusions The staff agrees with the licensee's evaluation of this modification. The license performed a comprehensive evaluation of this rodification, including safety and electrical analyses. Although this modification will not result in any change in safe +y margiris, the increased orifice valve speed may increase the frequency wherewith operators must take corrective action to comply with LCO 4.1.7 and LCO 4.1.9. On the basis of our review, we determine that this modification is within the scope of 10 CFR 50.59 and does not constitute an unreviewed safety question.

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3. _CN-1983 - Systems 21827-Primary Coolant System and Secondary Cooling System

1. Why Selected This change was selected for review because it involves modification to the plant protective system and affects the circulators, which are important for decay heat removal.

2. Sumary of the Change This change modifies the helium circulator control system to avoid the trip of a second circulator in a loop on overspeed following the trip of the first circulator. Past operating experience has indicated the overspeed condition occurs above a power level of 70%. The function of the helium circulators is unchanged, and their reliability is improved.

3. Tests for Significance This modification is safety related because it involves modification of Class 1 instrument racks and pressure controllers. It was determined not to be safety significant since the modifications involved do not affect the capability to prevent or mitigate the consequences of accidents described in the FSAR.

The modificaticn of the helium circulator control system was determined by the licensee not to be an unreviewed safety question on the basis of the following. This modification has not increased the probability of occurrence of an accident previously evaluated in the FSAR. This modification will avert unnecessary circulator trips on overspeed. Also, the addition of the new instruments will enhance circulator availability and operability. The design function of the helium circulators is unchanged and therefore the possibility of an accident different than any evaluated previously in the FSAR has not Deen created. Since new instruments will be procured and installed in accordance with applicable specifications, and auxiliary electric room instrument racks and power supplies were designed for additional loads, the margin of safety is not reduced.

This modification requires no changes to the applicable sections of the TS (LCO 4.2.1 and 4.2.2).

4. NRC Evaluation and Conclusions The staff agrees with the licensee's evaluation of this modification. The licensee has perfonned an evaluation of this modification, including safety, electrical, and structural analyses. On the basis of our review, we determine that this modification is within the scope of 10 CFR 50.59 and does not constitute an unreviewed safety ouestion.

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4. CN-1889 - System 11-Reactor Yessel and Internal Components

1. Why Selected The change involves modifications to the reactor coolant pressure s boundary. Since depressurization accidents are credible at Fort St.

Vrain, and potentially iinportant, this change was selected for review.

The change also involves' instrumentation used to maintain compliance with TS limits on reactor pressure.

2. Sumary of the Change This modification relocates several differential pressure instrumentation lines from within PCRV penetraticns C1 and C3'to outside the penetrations.

This will allow the replacement of the existing transducers with more stable, reliable, and accurate, but physically larger, transducers, which would not fit within the inteript of penetrations C1 and C3. The new transducers will permit the licensee to better quantitatively verify that the helium flow rates, and therefores core cooling, are within the limits of LCO 4.1.9 during low power operation. The replacement of the actual transducers will be done in a subsequent phase of this modification.

This modification also increase the accessibility of the instruments, thus facilitating ease of calibration and maintgnance.

3. Tests for Significance i s s This modification is safety related since it involves Class 1 equipment.

This modification was detemined not to-be' safety significant since it A does not affect the capability to prevent or' mitigate the consequences of accidents described in the FSAR.

The relocation of several differential pressure instrumentation lines from within the PCRV penetrations C1 and C3 to outside the penetrations was determined by the licensee not to be an unreviewed safety question ,

on the basis of the following analyses. The proposed perforation of each '

of the helium circulator instrumentation penetration vessel walls (tertiary boundary) with eight holes, through which instrumentation lines l will be run, will not degrade the tertiary pressure boundary provided by the penetration vessel. In the event that all eight instrument lines of ,

one of the helium circulator instrumentation penetrations were severed l between the inconel sleeve and the first isolation valve, the flow rate of helium would be twelve times less than that of the maximum credible accident evaluated in the FSAR. The failure of the pressure differential elements or instrumentation will not cause unsafe conditions and does not represent an accident of a different type than evaluated in the FSAR.

Because this modification will allow the core cooling limits of LCO 4.1.9

, to be more accurately monitored, the margin of safety associated with preventing flow stagnation in core refueling regions will not be decreased and safety will be enhanced. Also, this modification will oct affect the pressure of the interspaces between penetretion primary and secondary 4

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l Page 7 closures (LC0 4.2.7), nor will it affect the lea'k rates from the primary and secondary closures when operating under the conditions of LC0 4.2.7 and.,tC0 4.2.9.

This modification requires no changes to the applicable sections of the TS (LCO 4.1.9 (Core Region Temperature Rise)).

4. NRC Evaluation and Conclusions The staff agrees with the licensee's evaluation of this modification. The licensee has performed a comprehensive evaluation of this modification, including safety, electrical, and mechanical analyses. On the basis of our review, we determine that this modification is within the scope of 10 CFR 50.59 and does not constitute an unreviewed safety question.

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5. CN-1272 A, B - System 92 - Electric Power System

1. WW Selected

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The F?rt SC Vrain emergency electric power system has been the focus of several other licensing actions concerning the reliability of the overall system. This CN changes the way power is distributed through this system and potentially could affect system reliability.

2. Summary of the Change These changes provided power to two computer power distribution panels.

The computers reouired an uninterruptable source of power. This was provided by existing inverters drawing power from the DC battery buses.

Thus, new loads were added to the DC buses.

3. Tests for Significance

1. Clearly safety related in that these loads are on the DC bus (which is Class 1E).

2. Since the DC buses are part of the emergency electric power supply, additional loads on the DC bus potentially reduce the capability to supply loads required to prevent or mitigate accidents.

3. This modification does not involve an unreviewed safety question, except as follows: The battery capacity was originally analyzed and found to be adequate. Later, large batteries with larger margins were added. The additional DC loads added have been shown to be well within the capabilities of the batteries, so a large margin of safety remains, although slightly (not significantly) reduced.

4. NRC Evaluation The NRC staff agrees with ?"'s overall determination concerning CN-1272 A, B under 10 CFR 50.59. Howver, the PSC conclusion in the summary safety evaluation was unclear. It did not clearly conclude engineering had performed a careful reanalysis of the additional loads on the battery system.. The PSC reviewer did not note that although the battery capacity was not covered in the TS or FSAR, it probably should be so specified.

5. Cc"clusions PSC mR e the correct determination. However, the safety evaluation process was affected by the incomplete analysis in the FSAR and lack of a clear TS basis for battery capacity. Documentation in the CN did not clearly indicate why the change was acceptable.

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6. CN-1977 - System 92 - Electric Power System

1. Why Selected The Fort St. Vrain emergency electrical power system has been the focus of several licensing actions concerning the reliability of the overall system. Environmental conditions of the emergency diesel engines could potentially have an effect on the reliability of the engines. The water jacket heaters and their associated cabling can have an effect on these environmental conditions and, therefore, this change was selected for detailed review.

2. Sumary of the Change The existing heater power cables were removed and new cables were installed. The upgraded cables were adeouately sized and protected against harsh environmental conditions. The new cables met separation criteria and were placed in seismically qualified cable trays.

3. Tests for Significance

1. Clearly safety related Class IE items.

2. Since the emergency electric power supply is important to mitigate accidents with loss of offsite power, these cables have safety significance. If the cables fail, the standby reliability of the diesels will clearly be degraded.

3. This modification does not involve an unreviewed safety question.

The new installation is clearly better than the old one and the probability of a cable failure is reduced. Thus, the margins of safety are increased. Since the heaters are not specifically mentioned in the TS, no change to TS is involved. Even if an upgreded TS specifically had mentioned heater operability, this change would not have affected that TS.

4. NRC Evaluation The NRC staff agrees with the PSC evaluation. The PSC evaluation did not address in detail the lack of involvement relative to TS changes.

PSC's evaluation against the tests for an unrsviewed safety question did not reflect a good understanding of these tests.

5. Conclusions PSC made the correct determinations, but clearly the quality of the 10 CFR 50.59 process could be improved.

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7. CN-1979 - System 92 - Electric Power System

1. Why Selected Thi!, change invc1ved two major areas under NRC licensing review. The first is Fire Protection, where PSC is still making plant modifications to achieve compliance with Appendix R. The second area is Building 10, were PSC made ma.ior safety system modifications without prior NRC approval. Since this change fell into both areas, it was considered appropriate to review.

2. Summary of the Change The existing pcwer sources for certain fire detection and emergency lighting systems were distribution panels that were being modified with an interrupt switch for fire protection considerations. To provide a new uninterruptable source of power, these loads were being placed on an ur.interruptable instrument power bus.

3. Tests for Sionificance

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1. Clearly safety related items (Fire Protection).

2. Since the fire detection system and emergency lighting are important to detecting and mitigating fires, the potentially improved reliability of these systems has safety significance.

The modifications to these systems should improve their performance if the electrical circuits involved were not degraded (overloaded).

3. This change did not involve an unreviewed safety cuestion. The overall reliability of the systems in question was enhanced.

However, the potential overloading of the new power source was not thoroughly addressed. Also, while the modification was in progress, there was no compensatory measure taken to assure activation of the Halon system.

4. NRC Evaluation The NRC staff agrees with the PSC determination concerning CN-1979 under 10 CFR 50.59. However, PSC evaluation was considered incomplete in that the technical analysis of the new power source was not thorough. Potential panel overload was not considered. The analysis also failed to note that.

the Halon suppression system in LC0 4.10.2 was affected by this change, and it was not concluded that interim compensatory measures were needed to l activate the Halon system should a fire be detected.  !

5. Conclusions PSC made the correct overall determination. However, the analysis failed to determine that interim compensatory measures to activate the Halor, system were needed when the detectors were temporarily deenergized.

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8. CN-1996 - System 92 - Electric Power System

1. Why Selected This change involved adding additional potential loads to the essential buses (emergency electrical power syster). Although not directly involving a safety system, safety systems could potentially be degraded by the changes involved here. The change also potentially interfaced with safety system controls.

2. Sumary of the Change A new load, the testing eouipment for circuit breakers, will be added to the loads on Essential Bus No 2. This load is not safety related. To prevent it from overloading the diesels in an emergency situation, an automatic dropout feature was added to the circuits.

3. Tests for Significance

1. Safety related in that the power source is a Class IE system.

2. Since the emergency electric power systen is important to preventing or mitigating accidents, these changes could 1 potentially degrade this system and affect the probability '

and potential consequences of an accident.

3. This modification nominally does not involve any unreviewed safety question. However, PSC's basis for relying on specific eouipment involved to protect safety related equipment from being degraded was not clear. Only after questions by the NRC staff, was it established that safety grade equipment would be used in this protective feature. Thus, failure of the non-safety system would most likely not affect the safety system.

4. NRC Evaluation The NRC staff concludes that the PSC determination concerning CN-1996 under 10 CFR 50.59 was inadequate. The PSC evaluation was very incomplete in terms of evaluating the potential degradation of the safety system (Essential Bus No. 2) from failure of the dropout relay. Potential interactions with the contact from Relay XCR-0212-1 were not examined.

Potential failure of the dropout relay system was not clearly established as unlikely because this was Class IE equipment.

5. Conclusion PSC made the correct overall determination. However, PSC's inadequate evaluation in this case could have led to a potential system degradation in the emergency electric power system. The documentation in the CN did not clearly establish why the change was acceptable. This portion of the PSC 10 CFR 50.59 review process needs to be strengthened.

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9. CN-2038 - System 21 - Primary Coolant System

1. Why Selected This change involved the bearing water system. This system has known reliability problems as witnessed by LER 86-017. It also is a fairly complex system, but important to reliable operation of the helium circulators. Since it also supports safe shutdown cooling, it was selected for review.

2. Sumary of Change Three check valves at the outlets of the bearing water supply pumps were removed. Nominally, since there is no differential pressure across the pumps when they are idle, the check valves are not needed: Hcwever, valves may protect against failures of other system components.

3. Tests for Significance

1. Clearly safety related Class 1E system.

2. Since the bearing water system supports decay heat removal through the circulators, the system is important to preventing and mitigating accidents. In this case, the removal of these check valves may increase the vulnerability of the circulator system.

One example is increasing the potential for one check vdve failure to defeat the backup bearing water system by allowing backflow through the normal bearing water system. In further discussions, the PSC staff noted that other existing valves would perform this protective function.

4 NRC Evaluation The NRC staff concludes that PSC made the correct determination.

However, the review was not thorough in that all the potential roles of the check valves were not clearly considered in the review.

5. Conclusions PSC made the correct overall determination. However, the quality of the 10 CFR 50.59 review process was not adequate to support the determination in that all the possible consequences of the change were not thoroughly examined. This portion of the PSC 10 CFR 50.59 review process needs to be strengthened.

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Enclosure 2 PSC Staff in Attendance at December 16, 1986 Audit of Fort St. Vrain Change Notices M. Holmes J. Selan J. Johns C. Owens No written material was provided to PSC at the time of this audit.

l Audit results were also discussed with J. Gramling at the Fort St. Vrain Site on December 17, 1986.

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