ML20247R212
| ML20247R212 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 05/31/1989 |
| From: | Wilson R GENERAL PUBLIC UTILITIES CORP. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| 5000-89-1772, NUDOCS 8906070171 | |
| Download: ML20247R212 (43) | |
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GPU Nuelear Corporation f lu ps E
One Upper Pond Road Parsippany, New Jersey 07054 201-316-7000 TELEX 136-482 Writer's Direct Dial Number:
May 31, 1989 5000-89-1772 Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Document Control Desk
Dear Sir:
Subject:
Oyster Creek Nuclear Generating Station Docket 50-219 Safety System Outage Modification Inspection Response to Report 88-202 This letter forwards GPU Nuclear's (GPUN) response to concerns identified in Inspection Report 88-202, dated February 17, 1989, which provided results of the design portion of the Safety System Outage Modification Inspection (SSOMI).
Our response was requested within 60 days. Due to resource constraints associated with the recent refueling outage, we requested an additional month to prepare our response. The NRC Project Manager for Oyster Creek confirmed that this additional time was acceptable by telephone on April 17, 1989.
Attached are our responses to the specific concerns contained in Appendix C of the inspection report which are identified as open.
GPUN was requested, in previous correspondence dated November 16, 1988, to provide resolutions to five concerns, identified during the SSOMI, prior to plant restart from the recent refueling outage. CiJN submitted letters dated December 12, 1988 and January 19, 1989 and met witn members of the NRC staff on January 30, 1989 to address those concerns.
The staff's February 17, 1989 letter noted two areas of particular concern.
These are the control room heating, ventilation and air conditioning (HVAC) system and the loading condition of the emergency diesel generator (EDG) buses. Based on our review, we conclude that the control room HVAC system is adequate to perform its required function and that the commitments described in our letter dated June 4, 1985 are met.
8906070171 890531 Mf PDR ADOCK 05000219 Q PDC i g GPU Nuclear Corporation is a subsidiary of General Pubhc Utilities Corporation
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As regards.EDG bus loading, the loading profiles are within the capabilities of the' machines. Operator training has stressed the importance of maintaining EDG load'within rated capacity. We agree that better information could have been provided to operators for manually' loading and unloading the EDGs and have revised the operating procedure accordingly. The random loads are small and we believe they do not significantly affect the loading profile, however, we will further evaluate this, as discussed in the attachment. The administrative process for monitoring EDG load growth has proven effective. We will, however, make this process more formal.
We acknowledge the staff's identification of programmatic strengths in our engineering and design effort. . These are a result of'our continuing experiences with, and assessments of, design process controls and implemented changes based on our reviews.
lf y t uly yours, 5 ih R. F. Wi son Vice President Technical Functions RFW/crb Attachment cc: Mr. Gus C. Lainas, Acting Director Division of Reactor Projects I,II Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20E55 Ragional Administrator Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector Oyster Creek Nuclear Generating Station Mr. Alex Dromerick U.S. Nuclear Regulatory Commission Mail Station F1-137 Washington, DC 20555
s- s ATTACHMENT SAFETY SYSTEMS OUTAGE MODIFICATION INSPECTION INSPECTION REPORT 88-202 GPU NUCLEAR RESPONSES TO NRC CONCERNS l
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The SSOMI team identified various concerns regarding the control room HVAC system and other issues. The NRC statement of these concerns and GPUN responses follows NRC CONCERN
" Deficiency A-2: Temperatures in the control room may exceed allowable limits during a postulated LOCA concurrent with a Loop.
DISCUSSION:
During the Cycle 12R refueling outage, the licensee implemented the addition of a new independent HVAC system to the control room cnvelope. The intent of this modification was to fulfill GPUN commitments to requirements arising from NUREG-0737, Control Room Habitability. The modification (BA 402854P2) consisted of the addition of a new rooftop air-conditioning unit and supply and return duct work which interconnected with the existing system duct work. New isolation dampers, controls, and associated instrumentation and power circuits were also provided. Upon completion of these foodifications, the new independent HVAC system (System B) functioned as the lead system, and the existing HVAC system (System A) performed the same functions when required to meet' single failure requirements. The new HVAC system and components were classified as " Regulatory Required" and will be powered from the unit substation 1B3 associated with emergency diesel generator No. 2.
The team reviewed this modification in detail, including modification design descriptions, drawings, and a number of calculations performed to substantiate the design and operation of the system. The team identified several concerns, described below, related to the design of the control room HVAC system,
- a. Section 4.1.3 of modification design description MDD-OC-826B, Division II, stated that, in the event of a postulated LOCA coincident with a loss of offsite power, System B will be " rendered inoperative."
Emergency diesel generator 2 power to HVAC System B loads will be delayed for approximately 30 minutes. The team reviewed several calculations related to the capability of the control room HVAC system to maintain temperatures in the control room within allowable limits during and subsequent to this time. ;
(1) GPUN Calculation 1302-826-5360 was performed to evaluate the f maximum steady-state temperature in the control room and cable
! spreading room, assuming no cooling water to the cooling coils of I the previously existing HVAC unit. This condition corresponded to the worst case scenario of concurrent LOOP and LOCA followed by a 30-minute waiting period when only the HVAC unit fan would be j manually loaded onto the diesel. The team identified the :
following concerns about this calculation.
(i) The establishment of individual device heat loads used for electrical equipment in the control room was based on two Burns & Roe calculations and a referenced Burns & Roe design standard. However, the actual heat loads listed in ;
the Burns & Roe calculation were lees than those listed in the guidelines provided by the design standard. For l
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P:g) 2' example, the design standard-indicated heat loads of.10 watts per indicating light and 25 watts per relay or I
meter'. By contrast, the calculations used 7 watts per light, 7 watts per_ relay, and 2= watts per meter. There F was no justification given in the calculations for this
- departure.from the design standard. In addition, the design standard did not provide any basis for much of the equipment listed in the calculations, and the calculations did not provide any basis'for the heat loads assumed."
(Concern 1)
"(11) The calculations did not include a transmission load from an adjacent area at a higher temperature than the final calculated temperature in the control room." (concern 2)
"(111) Thero was no basis for the 800 Btu /hr heat load assumed for the occupants of the control room." (concern 3)
II "(iv) Transmission loads from areas adjacent to the lower cable spreading room at temperatures 135 degrees-F higher than-the final calculated temperature were not considered."
E (Concern 4)
"The team also reviewed calculation 15050-M4-003 performed to size the new 7ooling unit for System B. In contrast to the loads assumed in the above calculation, this analysis indicated substantially greater equipment heat loads (approximately 20 percent more) and provided a reasonable basis for these loads. In addition, the heat load assumed for the control room occupants was higher and was based on six people. [The team noted:that the FSAR stated that the normal operating capacity of the control room was seven people and the emergency capacity was'15 people. Burns &
Roe Calculation 10.000.09, discussed in paragraph a(2), below, assumed 10 people were in the control room under the same conditions.)" (concern 5)
"Recent air flow tests on the existing HVAC system (System A) indicated air flow rates which were 15 percent lower than the required 14,000-cfm design flow (see Inspection observation SMK-3). Thus, calculated temperatures which were based on the 14,000-cfm design flow rate will be further increased due to the
[ lower actual flows achievable when the existing cystem operated as a backup." (Concern 6)
"The final temperature calculated during the loss of offsite power with only the cooling unit fan operating was 104 degrees F. This I
was also the maximum allowable ambient temperature for standard rated electrical equipment, as stated in the FSAR. The team was concerned that this limit will be exceeded by a significant amount (upwards of several degrees) if the above concerns were considered in determining the control room temperature under these conditions. The licensee stated that the impact of potentially higher temperatures on the control room equipment for this case was not known. However, Control Room Operating Procedure No. 331 indicated that instrumentation may degrade at 90 degrees F and the computer may be damaged at 85 degrees F." (Concern 7)~
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"(2) Burns & Roe Calculation 10.000.09 is a room heatup calculation.
performed to determine the control room < temperatures.resulting from a loss of all HVAC. Results indicated that a temperature of 96 degrees Ficould be reached within one hour. However, this i calculation was based on the same heat loads questioned in paragraph a(1), above. In addition, there was no bas'.s given to '
substantiate the assumption that heat transfer coeffi 19nts;would-
.be equal on both sides of the control room walls, ceilings,.and
- floors.
The team was concerned that control room temperatures for the time period prior to loading the diesel with the HVAC. fan (only) may be higher than indicated by these calculations and may exceed the 104 degrees F equipment limits."'(Concern 8)
"b. Section 4.1 of modification design description MDD-OC-826B, Division I, indicated.the design ambient conditions as 89 degrees F (high) and 10-f- degrees F.(low), based on the OCNGS FSAR. These temperatures were not I- expected to be exceeded more than 2.5 percent of the time, and they reflected the original design basis for the plant. The team identified the following in.this MDD.
(1) These temperature extremes and their frequency of occurrence may not be an adequate design basis for a modification made'today to a critical plant HVAC system such as that for the control room.
(2) Even accepting these temperatures and their frequency of occurrence as an adequate basis for sizing the control room HVAC system, their use was not conservative as a maximum outdoor temperature to. determine control room steady-state temperatures
!~ with fan operation only.
The combined effects of the nonconservative heat loads assumed in the analyses, measured' air flows substantially less than required design flows, and-the potential for high ambient temperatures exceeding original plant design basis temperatures could result in temperatures which exceed control room equipment capabilities."
(concern 9) i "Both the existing and the new control room HVAC systems are not safety-related. However, this observation has some safety significance since. l the modification to add a new control room HVAC system represents GPUN's response to commitments arising from NUREG-0737. The team had the following concerns.
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- a. These commitments may not be satisfied in view of the issues raised in this observation,
- b. The capability of electrical equipment and instrumentation in the control room required to monitor the course of an accident may be compromised during a coincident loss of offsite power.
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l? sin addition, the team was unable to determine when the full control room HVAC system equipment could be loaded onto the. diesel. The licensee stated that i this was ' indeterminate' since it was based on operator discretion for the case-considered.and the prevailing circumstances. (See Inspection Observation C-1
- for further details on diesel generator loading and its relationship to control j
. room.HVAC.)1 Consequently,.it was not clear that adequate control room temperatures'will be maintained during a loss of offsite power" (concern 10)
"In response to the team's concerns, the licensee stated that the heat loads and calculations in question were being reviewed and evaluated. In addition, further testing may be conducted to establish actual heat loads during operation. This issue remains open." (concern 11)
" REGULATORY BASIS:
FSAR Section 9.4.2.1 states that plant HVAC systems are designed to limit temperatures so that the ' maximum allowable ambient temperature for standard rated electrical equipment (104 degrees F) is not exceeded.'"
GPUN responses to the concerns identified above supplement our previous correspondence (GPUN letter 5000-88-1685 dated December 12, 1988) and are as
.follows:
GPUN Response to Concern 1 This concern was addressed in our letter (5000-88-1685) dated December 12, 1988
" Response to Restart Concerns." See Item 3 of that letter.
GPUN Response to Concern 2 Calculation 1302-826-5360-001 has been revised to include transmission loads from adjacent areas which are expected to be at a higher temperature than the final calculated temperature in the control room. (See response to Concern 9.)
GPUN Response to Concern 3 Calculation 1302-826-5360-001 has been revised to delete the 800 Btu /hr heat load previously assumed for control room occupants. The new heat load for occupants is in agreement with ASHRAE Fundamentals, 1981, and is 315 Btu /hr/ person. This data is consistent with the calculations for sizing the new System B A/C unit.
GPUN Response to Concern 4 Calculation 1302-826-5360-001 has been revised to take into account transmission heat loads from areas adjacent to the lower cable spreading room, which are expected to be at temperatures higher than the final calculated room temperature. (See response to Concern 9.)
Paga 5 GPUN Response to Concern 5 As discussed in Item 3 of GPUN Letter 5000-88-1685, dated December- 12, 1988, the electrical heat load calculation (15050-M4-003) is conservative. The heat
- load was developed using a certain wattage per square foot of front panel area instead of individual component losses. The new revision of Calculation (3731-29-E-004, dated December 5, 1986) provides a more accurate accounting for-controlLroom heat losses.- This calculation confirms the conservatism of the System B sizing calculation.
Calculation 1302-826-5360-001 has been revised to include control room occupancy of 15 persons per FSAR eniergency condition discussions. B&R calculation 10.000.09 (Rev. 2), which determines control room steady state temperature following a LOCA with LOOP event, still assumes 10 person occupancy and not 15 people, as stated in the FSAR. The impact of the additional five persons, as a heat load, on the control room temperature is considered negligible.- The additional heat load is 1575 Btu /hr (five persons x 315 Btu /hr/ person) which is 1.3% of the total control room heat load of 123,600 Btu /Hr. Calculation 15050-M4-003 still assumes six person control room occupancyLand has not been revised. The impact of an additional 9 person heat load on the A/C unit is negligible and is compensated by the conservative electrical heat load estimate.
GPUN Response to Concern 6 The impact of HVAC System A reduced air flow on control room temperature is addressed under Deficiency'A-3.
GPUN Response to Concern 7 See response to Item 3 of GPUN Letter 5000-88-1685, dated December 12, 1988, and additional information as supplemented below:
Operating Procedure'No. 331 has been revised to delete the warnings related to degradation of instrumentation at 90'F/85'F. In the current procedure, operators have been warned that " instrument reliability may degrade as room temperature reaches 104 F".
GPUN Response to Concern 8 See response to Item 3 of CPUN Letter 5000-88-1685, dated December 12, 1988, and additional information as supplemented below:
The concern related to the non-substantiated assumption that the heat transfer coefficients would be equal on both sides of the control room walls, ceilings and floors is minor. Per ASHRAE-Fundamentals 1981 for the "still air" ]
condition which is the case for all ceilings, floors and walls except one exterior wall, the assumption is valid. For one exterior wall, if an outside wind condition of 15 mph is assumed, the film coefficient is different, R=0.17.
instead of 0.68 as considered in the calculation. The impact of this film coefficient difference on the overall heat load due to additional heat
- . transmission is negligible, less that 0.1%.
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-Pags 6 GPUN Response to Concern 9 The use of 89'F/10*F (2.5% value outdoor temperature) as an adequate design temperature is in accordance with the recommendation of ASHRAE - Handbook of Fundamentals for massive buildings with little glass. The Oyster Creek control room is typical of this description. A massive structure will reduce the effect of overload from short intervals of outdoor temperatures above the design value. The NRC concern, which considers accepting these temperatures and their frequencies as an acceptable outdoor design condition, is more related to the use of these temperatures as a maximum temperature to determine Control Room steady state temperature during LOCA/ LOOP with fan operation only.
Our previous response, Item 3 of GPUN Letter 5000-88-1685, dated December 12, 1988, indicated that following a LOCA/ LOOP, the ventilation system is shut off. Based on the revised calculation of heat load with half of the control room lighting off (which is the normal condition), the maximum allowable room temperature of lO4*F is expected to be reached in approximately three hours.
Considering restoration of power to control room equipment within two hours, as discussed in our previous letter, the need to run the supply fan without cooling to bring in outside air is neghted. Within this time, if outdoor temperature permits, the fan may be used to bring in 100% outdoor air to cool the room. The maximum temperature limit to use outside air to cool the control room has been revised to 82*F from 89'F as previously indicated. The calculation prepared to determine the maximum temperature of the outdoor air which can be used during " fan only" mode of operation, uses the revised heat load, reduced air flow rates as measured during recent tests, and includes the transmission losses from adjacent areas which are expected to be at a higher temperature. In conclusion, by limiting the outdoor temperature conditions at 82*F during LOCA/ LOOP (fan mode of operation only) the concern related to room temperature exceeding equipment capabilities is negated. Operating procedures l have been revised to incorporate provisions te instruct the operator to limit the use of outside air to cool the control roon following LOCA/ LOOP to only when the temperature is less than 82*F. Therefore, following LOCA/ LOOP, the 104*F control room temperature limit is protected and capabilities of
' electrical equipment and instrumentation in the control room to operate during the course of an accident are not compromised.
GPUN Response to Concern 10 l The time when an entire control room HVAC system (including the cooler) can be l loaded onto the diesel generator is based on operator determination for the I
case considered and prevailing circumstances. Operators are trained, and the operating procedure has cautions, about the allowable temperature limits of 104*F and the time involved to reach this temperature. We expect that a complete HVAC system can be loaded onto the diese) within a one hour period.
i As stated in previous correspondence (December 12, 1988), other sources of I
power will also be available.
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'GPUN Response to Concern 11 j E
The following calculations have been reviewed and revised:
A. Calculation 1302-826-5360-001, control Room Loss of Chilled Cooling Water, has been revised to includs:
Provisions for 15 person control room occupancy and the basis for heat relaase/ person.
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Transmission loads from areas at higher temperature than the control room final temperature.
New electrical heat load per the latest revision of Calculation 3731-29-E-004.
Reduced' air flow of System A per latest test results, dated February 3,.1989.
B. Calculation 3731-29-E-004, Heat Load Estimate for the Control. Room, has been revised to include:
The results of the system walkdown of actual electrical components employed in the control room, manufacturer's literature for
' documentation of actlal heat losses being released to the control room environment, B&R Engineering Standard - Heat Losses from Electrical Equipment or conservative engineering judgment, as identified.
C. Calculation 10.000.09, control Room Temperature Study With Loss of Ventilation, has been revised to include the new heat load released by electrical components as indicated in the latest revision of Calculation 3731-29-E-004.
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i PIga 8 NRC CONCERN
" Deficiency A-3: The existing control room HVAC system (System A) air flow capacity does not meet established design requirements.
DISCUSSION:
In conjunction with the review of documentation supporting the addition of a new independent HVAC system to the control room envelope, the team reviewed GPUN memorandum MSS-85-383. This memorandum dated July 29, 1985, discussed control room habitability differential pressure and air flow test results. The memorandum indicated that measured air flow rates for the existing HVAC system at this time were 11,400 cfm. This was substantially less than the present required design flow rates of 14,000 cfm. Some reaasons given for the low flows neasured include measurement errors and duct or damper leakage.
The licensee stated that more recent tests (TP200/0.1, Revision 0, MTX 26.11.5.5, March 23, 1988) indicated air flow rates of 11,909 cfm after maintenance work on the ducts anu dampers was completed. This flow was 15 percent less than the 14,000 cfm design flow rate for the HVAC system. Thie deviation exceeded standard industry practice limitations of 10 percent of design flow rates, including those of ANSI N510, " Testing of Nuclear Air-Cleaning Systems," 1980. In addition, calculations performed to establish control room temperatures were based on required design flow rates of 14,000 cfm. Lower actual flow rates will increase these temperatures.
The control room HVAC system is not safety-related. However, this observation has safety significance. The modification to add a new independent HVAC system was performed to meet guidance arising from NUREG-0737. The team was concerned that, assum.ng s a single failure of the new HVAC system, (lead System B), the capacity of the existing control room HVAC system (System A) may not be adequate to meet design requirements during emergency conditions. Further, since test results were based on the existing system's duct work, which was shared by the new system, there was no evidence that the new system will be capable of supplying adequate air flow rates.
REGULATORY BASIS:
10 CFR 50 Appendix B Paragraph III " Design Control" states in part that measures shall be established for the control of design interfaces.
Modification Design Description Division II (Preliminary), MDD-OC-826B, states that the control room HVAC system is capable of providing 14,000 cfm of outside air during a LOCA or main steam line break without exceeding maximum allowable radiological dose rates. The results of a number of design analyses (See Deficiency A-2 Observation for references) are based on the design flow rate of 14,000 cfm; however, the measured flow rate is substantially less tnan 14,000 cfm. This issue remains open."
GPUN Response:
The design air flow requirements to satisfy the cooling needs for the control room and cable spreading room is 13,500 cfm per flow diagram, Updated FSAR Figure 9.4-1, " Flow Diagram HVAC Office Building, Control and Cable Rooms," and not 14,000 cfm.
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... ... 1 Pag 2.9 The above-referenced air flow rate of 14,000 cfm represents the design fan capacity required initially to meet the cooling needs for the control room, cable spreading room and the observation room. The observation room (which required 500 cfm) was subsequently provided with an independent air conditioning unit which reduced the air flow requirements for the control room and cable spreading room from the initial design of 14,000 cfm to 13,500 cfm.
On February 3, 1989, a new control building ventilation upgrade functional test was performed (TP 254/11 MTX 26.12.1). 'The test was performed on both A and B systems.
For System A, the measured air flow rate for the emergency mode of operation was 11,494 cfm, which is less than the design requirement of 13,500 cfm by approximately 15%. Usually, a lower actual flow rate will increase room temperature. During a selected design' day of 89"F, a slight increase in room temperature is expected if the heat released in the rooms is as originally estimated. A revised calculation demonstrated that the heat released in the control room with the normal 50% lighting condition is approximately 20% less than originally calculated (see GPUN Letter 5000-88-1685, dated December 12, 1988. With this reduction in control room heat load, the reduced air flow rate does not have any impact on control room temperature. I Per the February 3, 1989 test results, the new installed System B is capable of supplying an adequate air flow rate, as measured, of 14,000 cfm.
Pago 10 NRC CONCERN
" Deficiency A-5: Control room operating procedure directions to open doors on loss of HVAC have not been evaluated for effects on operating personnel resulting from a potential coincident radiological incident.
DISCUSSION:
Inspection Observation A-2 discussed inadequacies identified in design analyses related to the modification which added a new control room HVAC system to the control room envelope. This observation determined that control room temperatures may exceed allowable limits in the event of a loss of offsite power, resulting in a loss of control room HVAC. During this review, the team noted that Operating Procedure 331 provided instructions to 'open the three doors to the Control Room and the door to the old cable spreading room to allow air flow from the corridors' in the event of a loss of control room ventilation.
l The team questioned whether this condition had been evaluated or considered in the analysis of radiation dose to control room operators should a coincident radiological event occur on a loss of offsite power. The licensee stated that the effects of airborne radiation or radiological exposure to operating personnel for this condition had not been considered. However, GPUN was evaluating this case.
The team reviewed the results of Stone & Webster Calculation No. 006, Revision 1, ' Post-LOCA Gamma & Beta Doses in the Control Room vs. Outside Air Intake Rate', July 27,.1988. Although the analysis did not specifically refer to the conditions described in this observation (open control room doors), results i indicated that significant margin existed in meeting required dose limits.
Consequently, this observation is probably not cafety significant. However, this item remains open pending formal resolution by GPUN.
REGULATORY BASIS:
10 CFR 50, Appendix B requires in Paragraph III, ' Design Control' that measures be established to transfer requirements end design basis currently to specifications, procedures drawings and instructions. Stone & Webster Calculation 006 invokes radiation dose limits for control room habitability based on NUREG-0800, Section 6.4. These limits were not reflected in OCNGS j Procedure No. 331, Revision 15. This issue remains open."
GPUN Response: j As indicated in previous correspondence (Item 3 of GPUN Letter 5000-88-1685 l dated December 12, 1988), Procedure 331 has been revised to instruct operators not to leave control room doors open during a radiological release. ,
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P ga 11 NRC CONCERN
" Deficiency A-6: No documented basis for safety-related motor-operated. valve maximum differential pressure and valve stroke time acceptance criteria.
DISCUSSION:
i A modification to a number of safety-related motor-operated valves (MOVs) consisted of relocating the open indication light contacts from Switch No. 7, Rotor No. 2, to another switch and rotor within the valve operator. The relocated open indication light switch was adjusted to turn off when the valve was 97 percent to 99 percent closed. The modification was performed in response to INPO SOER No. 86-2 to reduce inaccuracy in MOV position indication and to reduce personnel error and incorrect position indication. l The team reviewed this modification and had no issues with the changes being made. However, the team identified weaknesses in documentation substantiating the design basis for these MOVs related to design differential pressures and acceptance criteria for valve stroke times.
OCNGS Procedure A100-GME-3918.51 provided instructions for testing Limitorque MOVs using Motor-Operated Valve Analyses and Test Systems (MOVATS). As part of this testing, valve stroke time was measured. Acceptance criteria for valve stroke time were based on previously recorded stroke times. In some cases, such as for the main steam isolation valves and containment isolation valves, the acceptable stroke times were based on requirements established in the FSAR.
However, only the main steam isolation valve stroke time could be traced to a documented design basis. The stroke time for this valve was based on a requirement established in a General Electric Specification. Acceptable stroke times for containment isolation valves were established based on the FSAR assumption that significant fission product release to the containment atmosphere was on the order of minutes for the design basis loss of coolant accident. However, this evaluation was not documented. In addition, there was no documented basic to ensure that the stroke times for all other safety-related MOVs were acceptable. These valves may have required opening and closing times which were consistent with system or process requirements and may have been more stringent than those currently used.
In a letter to the NRC responding to IE Bulletin 85-03, GPUN noted that the bulletin did not apply to OCNGS systems and that a program was being implemented to encompass most of the bulletin's requirements. The NRC subsequently accepted further expansions of this program to include the core spray and isolation condenser systems. GPUN provided a revised response that included maximum differential pressure values for safety-related MOVs in the core spray and isolation condenser systems. The basis for maximum differential pressures provided in the response was limited to a brief description of the conditions which establish the maximum pressures. However, the licensee was unable to provide documented analyses to substantiate these maximum differential pressures. The team was unable to determine, from the brief descriptions provided in the response, whether worst-case conditions had been evaluated assuming a single active failure of any system component. The team noted that the table upon which the response was based was included in GPUN Calculation C-1302-900-5360-005. However, this calculation merely provided a calculation cover sheet for the table and contained no formal analysis, evaluation, or development of the results presented in the table. There was no basis provided for any parameters or data used in the table to establish maximum differential pressures.
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This observation could have some safety significance.1 There was no documented evidence;to ensure that maximum differential pressures have been' established ~
forfsafety-related MOVs in accordance with the requirements of IE Bullcsin 85-03 (such as-consideration of worst-case conditions and assuming a single active failure). This could affect torque switch settings for these valves.
Thus, there was no assurance that these valves will be capable of opening or closing against worst-case differential pressures during postulated accidents.
In addition,. valve stroke time' acceptance criteria applied during MOVATS testing may not be appropriate for some safety-related valves.
The licensee informed the team that substantiation of these design basis data for MOVs was a continuing effort at GPUN as part of the design basis L reconstitution effort. This item remains open pending resolution by the g licensee.
REGULATORY BASIS:
ANSI N45.2.11 requires design activities to be documented and the final design to be traceable to the source of design input."
GPUN Response A motor-operated-valve (MOV) differential pressure calculation was completed by GPUN Calculation No. Cl302-900-5360-005. This calculation included the <
following safety-related systems and valves:
TOTAL SYSTEM CIV OTHER MOV'S NUMBER OF VALVES Main Steam Drain 4 -- 4 RBCCW 3 1 4 Iso Condenser -- 8 8 RWCU 4 7 11 Shutdown Cooling 2 6 8 Core Spray -- 12 12 Containment Spray -- 10 10 57 Total calculation C1302-900-5360-005 does provide bases for the maximum differential pressure for motor-operated-valves. However, after reviewing the calculation, we believe the way this calculation is presented could be improved. More
' descriptive comments and additional references will be added. Further, we believe that the calculation does conrider the worst process condition for valve delta P (including single fai*ure). We will review the calculation to confirm this.
A valve stroke time criteria document does not exist to our knowledge. To establish such a document for all safety-related MOVs would take considerable effort. However, to address the specific concern, we are taking the following approach to develop a basis for safety-related motor operated valve stroke times:
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- l. 'For containment isolation valves, stroke time acceptance criteria per Technical Specification 4.5.I, and~its basis will be used.
- 2. .For the' Core Spray. System, the stroke time basis has been recently' :)
developed (Reference GPUN letter to the NRC dated 3/15/89.) j
- 3) ,For the Containment Spray System, a design basis reconstitution effort-
! is underway and scheduled to be completed by December 1989. The valve stroke time basis will be part of this effort.
- 4) For the remaining systems,. justification will be provided for the-existing stroke time acceptance criteria.
Considering the scope of the above work we expect to complete this work by December 31, 1989.
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NRC CONCERN
" Deficiency B-1 Instrument setpoints, as specified, do not meet design criteria.
DISCUSSION:
This modification (BA 402896) replaced several sets of mechanical pressure f switches with analog loops to monitor reactor vessel pressure. Analog loops ]
d RE03A through RE03D provided inputs to the Reactor Protection System (RPS) relay logic to scram the reactor when their pressure setpoints were exceeded.
Similarly, analog loops RE15A through RE15D provided inputs to the Engineered Safeguards Actuation System (ESAS) to trip the recirculation pumps and initiate i isolation condenser operation when their pressure setpoints were Exceeded.
The team found that the reactor pressure setpoints as specified for the new analog loops did not meet the design criteria. Section 4.8.3.3 of modification design description MDD-OC-622A, Division I, specified that the REO3 analog loop for reactor vessel pressure should actuate before the RE15 analog loop. In addition, Safety Evaluation SE-402896-002, Revision A, stated in Section 3.3.1 that the existing condition where the RE03 setpoints are set below the REIS setpoints is continued in the new analog loops. The new setpoints for the RE03 ,
and RE15 analog loops were determined in calculation 4283-12-11-001. The team I found that, under certain conditions, the RE15 analog loop would actuate before the RE03 analog loop due to overlapping of the setpoints. This would occur when the RE15 analog loop had a negative 1 cop accuracy error and the RE03 analog loop had a positive loop accuracy error.
A trip of the recirculation punps and initiation of the isolation condenser operation before a reactor trip would put the plant in an unanalyzed condition. The licensee agreed to reevaluate the calculation of the setpoints and specify new setpoints which would eliminate setpoint overlap.
REGULATORY BASIS:
i ANSI N45.2.11, Quality Assurance Requirements for Nuclear Power Plants, Section 6, Design Verification, requires that calculations to be properly reviewed to ensure that the design criteria is met."
GPUN Response The above concern was addressed in Item 1 of GPUN Letter 5000-88-1685 dated December 12, 1988.
P:ga-15
!l HEQ CONCERN 1" Deficiency B-2 L Technical specifications were not revised to reflect an upgraded accident monitoring system as specified in safety. evaluation.
DISCUSSION:
-Modification BA 402256 installed a new suppression pool temperature monitoring system to permit more accurate calculation of suppression pool bulk.
temperature. Ths new system also provides the operator with improved capability;to monitor containment integrity. It was designed to meet the guidance in NUREG-0661, Appendix A'and USNRC Regulatory Guide 1.97, Revision
- 2. The suppression pool temperature was designated as a Regulatory Guide 1.97, Type A variable for DCNGS>
GPUN Safety Evaluation, SE 402256-003, Revision 0, Suppression Pool Temperature Monitoring System, Sections 2.6 and 2.7 stated that an amendment w s required to plant Technical Specification Sections 3.13 and 3.14, which apply to accident raonitoring instrumentation operating status and surveillance requirements. GPUN action item request AT 5123 was assigned to and accepted by the Oyster Creek licensing group with an expected completion date of February 28, 1989.
The team was told that an update of the Technical Specification will not be made to reflect this modification, pending resolution of issues regarding requirements for listing Regulatory Guide 1.97, Type A variables in the technical specifications. This is an item of discussion between the Boiling Water Reactor Owners Group (BWROG) and the Office of Nuclear Reactor Regulation.
Although there was no immediate safety significance, the team felt that plant safety could be enhanced by taking steps to ensure operability of these instruments.
The team was also concerned that the safety evaluation, SE 402256-003, Revision I
.0, continued to reflect that the plant Technical Specification, Sections 3.13 and 3.14 had been updated.
REGULATORY BASI}:
10 CFR 50, Appendix B requires that measures be established to transfer design basis to specifications, drawings, and instructions."
GPUN Response Based on discussions between the industry working group on technical specification improvements and the NRC Staff, it is our understanding that technical specifications for Regulatory Guide 1.97, Category 1, Type A monitoring instrumentation will not be required.
Surveillance requirements for the suppression pool temperature monitoring system are incorporated in plant procedures that meet or exceed the requirements established in the current BWR Standard Technical Specifications (NUREG 0123, Revision 3) for this system. In addition, Oyster Creek Technical Specification 4.5.P.1 currently requires a daily check of suppression pool temperature.
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Safety. Evaluation-SE402256-003, Revision'O did not indicate that Technical i
' specification sections 3.13 and 4.13 had been changed but that they would be changed.' The SE will'be revised to delete statements regarding changes to.'
Technical Specification sections 3.13 and 4.13.since no changes-are deemed
-necessary.
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P g) 17 NRC CONCERN
" Deficiency C-1 Insufficient guidance to reactor operators during a postulated LOCA and coincident LOOP could potentially overload the diesel generator.
DISCUSSION:
A review of the licensee's diesel generator load calculation (C-1302-741-5350-001, Revision 1) and Emergency Diesel Generator Operation Procedure No. 341, Revision 26 showed that, during a LOCA concurrent with a LOOP condition, the diesel generator had the potential of being overloaded.
This overload condition could be caused by the addition of manual loads (e.g.,
air compressors, battery charger, fire end pump, or control room HVAC and other ventilation systems) during an accident condition. The calculated worst-case i diesel loading (2703kW) due to automatically started loads demonstrated that I the diesel was of adequate capacity, but had only a small margin before it would exceed its overload rating of 2750 KW.
The engineering staff contended that manual loads were not'added to the diesel during the initial phase (first half-hour) of the postulated accident.
However, the Emergency Diesel Generator Operation Procedure instructed the operator to add manual londs to the diesel after automatic load sequencing was completed. The team was concerned that the operators were relied on to take appropriate action to mai.ntain acceptable diesel loading without suf ficient engineering guidance concerning manual loads that could be added or automatically started loads that could be secured. This lack of guidance could cause the diesel to become overloaded and possibly degrade the voltage of the class 1 buses and prevent safety-related equipment from performing its intended >
functions.
The license agreed to revise the Emergency Diesel Generator Operating Procedure to show a list of loads that may be removed to avoid overloading the diesel generator. This item remains open.
REGULATOR BASIS:
Appendix B to 10 CFR 50 states in part that measures shall be established to assure that the applicable regulatory requirements and design bases are correctly translated into procedures."
GPUN Response Oyster Creek Procedure 341, " Emergency Diesel Generator Operation" was revised as committed in GPUN Letter 5000-88-1685, dated December 12, 1988 (Item 5).
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.'P*g3 18 1
NRC CONCERN
" Deficiency C-2's overloading.the diesel during surveillance testing. i
. DISCUSSION:
~
A review of the diesel generator load test (Surveillance Testing Procedure No.
636.4.003, Revision 31, Serial Nos. 1036 and 1040) showed that.the diesel generator was loaded above its 10 percent overload rating of 2750 KW during surveillance testing. The surveillance testing was used to demonstrate that the diesel can operate at'a load level associated with emergency. standby design.
The calculated diesel. load was 2703 KW. The surveillance procedure required the diesel to be tested 50 KW above the overload rating, i.e., 2800.KW. The acceptance criteria requirement (Item 7.0 No. 2 of the procedure) specified any load _over 2700 KW (for one hour) with no restriction on maximum loading.
Written correspondence (letter from General Motors dated June 26, 1985) from the manufacturer stated that GPUN must bear the responsibility for operation above the 2750-KW rating.
The team was concerned that exceeding the manufacturer's maximum recommended overload rating will degrade the diesel generator and raised questions about its operability and ability to mitigate an accident. This item is open.
REGULATORY--BASIS:
The manufacturer recommendations that the diesel loading.should not exceed the 10 percent overload rating was not accounted for in Surveillance Testing Procedure No. 636.4 003 Rev. 31, Appendix B to 10 CFR 50 which states in part that measures shall be in place to assure design bases are correctly-translated into. procedures."
L GPUN Response The above concern was addressed in GPUN Letter 5000-88-1685, dated December 12, 1988 (Item 4).
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- Pags 19 NRC CONCERN
" Deficiency D-It Pipe Support and Supplementary Steel Stress Allowables DISCUSSION:
As part of GPUN Budget Activity 402876P3, the licensee separately qualified 11.
nuclear safety-related (NSR) piping systems to the Housner seismic response
. spectra (which form the current seismic licensing basis for OCNGS) and to the new Blume response spectra and corollary ASME code cases. GPUN provided the l- ~
NRC with a number of the details of its program to requalify NSR piping to new seismic criteria through meetings and correspondence. However, the team.could not confirm that.GPUN provided the NRC with details of the relaxed stress allowables which GPUN used to requalify pipe supports and supplementary steel when the pipe support loads predicted by the new Slume response spectra exceeded the pipe support loads predicted by the Housner. response spectra of
. record.
The piping code of record for DCNGS, ANSI B31.1-1983, specified an upset allowable stress factor of 1.2 for pipe supports and an upset allowable stress factor of 1.33 for supplementary steel (by reference to the AISC Code).
Revision 0 of the GPUN piping specification (Reference B) reiterated.these upset allowable stress factors. However, GPUN revised the piping. specification to permit the use of the Service Level D stress limit factor of 2.0, which was specified in the ASME Code (Reference C). This represented an increas'e of 67 percent in the allowable upset stress for pipe supports and an increase of 50 percent in the allowable upset stress for supplementary steel with respect to the allowable upset stresses which ANSI B31.1-1983, the piping code of record for OCNGS, permitted for these components.
GPUN licensing did not provide "Se team with a formal response to this issue during the week of October 31, 1988. GPUN informally confirmed the use of higher upset stress allowables to requalify some pipe supports and supplementary steel. This item remains open.
REGULATORY BASIS The upset stress allowables which GPUN is using to requalify some pipe supports and supplementary steel exceed the allowables permitted by the piping code of record'for OCNGS (B.31.1-1983) as specified in the FSAR." l GPUN Response I GPUN has used the Service Level 'D' allowable only in conjunction with 1987 new Blume response spectra curves. The application of the new response spectra for the Bulletin 79-02/14 upgrade program is discussed in GPUN letter 5000-88-1637, dated September 19, 1988. We have evaluated use of the Service Level D allowable with the new response spectra and found it to be acceptable. The new response spectra analysis accounts for multi-mode effects and in general results in higher calculated seismic reaction loads on pipe supports than were determined by the original design methods.
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.Pags 20 l
NRC CONCERN ]
'" Deficiency D-2: U-Bolt and Lever Arm Calculation i
DISCUSSION:
GFDN Budget Activity 328195P2 replaced internal and external parts for 14' '
' torus-to-drywell vacuum breakers to meet NRC Mark 1 containment program-
! acceptable stress criteria. A GPUN vendor-(Reference A) qualified the valve replacement parts to the Mark 1 program requirements. GPUN qualified the ,
U-bolt and lever arm configuration that is attached to each valve' shaft. These i
. configurations initiate an alarm in the control room on valve slamming ,
(Reference B). The U-bolt and lever arm configuration was shown schematically. j on the valve drawing'(Reference C) and in detail on page 2 of the applicable calculation (Reference B).
1 The team reviewed GPUN's calculation to qualify the sensing arm connection to l the design torque specified in Reference A, Appendix 1. .The team found the l Reference.B calculation deficient, in part, because of the following factors:
- a. The magnitude of the bending stress allowable for the U-bolt material .
was incorrect. The computed bending stress was compared directly to the !
material yield stress instead of the lesser allowable stress. l 1
- b. Tha tension stress in the U-bolt due to the fact that static. pre-load ;
and applied torque were not combined with the computed bending stress. ]
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- c. The calculation did not qualify the lever arm to the design torque.
GPUN licensing did not provide the team with a formal resronse to this issue a during the week of October 31, 198' 'ver, GPUN informally indicated that the calculation will be revised.
The team did not consider this issue to be safety-significant, since the sensing arm will activ3te the control room alarm on initial valve actuation, even if the lever arm is overstressed. This item is open pending revision of ,
the calculation.
REGULATORY BASIS:
The calculation does not meet the intent of Section 4.9.5.f, Mechanical j Requirements, of the Referenced GPUN specification, which requires that system-specific mechanical functional and design requirements be identified and satisfied. Appendix B to 10 CFR 50 states in part that these measures will be in place to ensure design bases are correctly translated into procedures".
EEUN Response GPUN concluded that the torus-drywell vacuum breaker position indication system ]
is not required for the vacuum breakers to perform their safety functions.
GPUN Calculation No. C-1302-243-5320-039 Rev. O, "18 Inch Torus to Drywell Vacuum Breaker Disk Positioning Arm", Jan. 14, 1988 has been voided.
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l Paga 21 NRC CONCERN
" Deficiency D-3: Seismic Qualification of Buried Emergency Service Water (ESW)
Pipe DISCUSSION:
GPUN Budget Activity 328141P2 was part of a long-term GPUN program to monitor pipe wall thickness in ESW piping following permanent removal of the corrosion-protective lining from portions of the piping. As noted schematically in Figure 3 of the GPUN report (Reference A), portions of the ESW lines were buried. The team asked GPUN to provide the seismic qualification report for these buried lines. The buried ESW piping was shown in detail on the related drawing (Reference B).
On November 3, 1988, GPUN provided the team with the seismic reanalysis report j (Reference 3) that summarized the calculation that seismically qualified the buried ESW piping. Through review of this report, the team noted that the piping material stress allowable used in the report (48 ksi) was based upon the use of ASME Code (Reference d) allowables, and was more than three times the upset allowable which the piping code of record for OCNGS permitted for Grade A-53 carbon steel pipe (1.2 x 12Ksi = 14.4 kei). Moreover, a review of the total stresses (axial & bending) summarized at eight locations for each of the two L-shaped segments of 14-inch diameter ESW pipe indicated that the piping consistently exceeded the permissible upset stress of 14.4 ksi. The total stresses summarized in the report varied from a low of 14.56 ksi to a high of 62.52 kei. ,
l On November 7, 1988, the team notified GPUN licensing of the above finding. At that time, GPUN indicated that the seismic reanalysis report had been submitted to the NRC for review, but that GPUN has not received a formal response to the submittal.
The team considered this unresolved item to be safety-significant, since the NSR ESW piping is required to perform a safety function during and after a safe shutdown earthquake (SSE). This item remains open.
REGULATORY BASIS:
FSAR.Section 3.7.3.2, Analytical Procedures for Piping, notes that
'All Class 1 (seismic) piping system configurations must satisfy the design stress requirements and allowables specified by ANSI B31.1-1983 Edition through Winter 1984 Addenda.'
Appendix B to 10 CFR requires, in part, that measures be estableihed to correctly transfer design bases to procedures, specifications, drawings, and instructions. The design bases specified in the FSAR were not correctly incorporated in the piping specification for OCNGS."
GPUN Response
Reference:
Seismic Reanalysis of the Buried Emergency Service Water Line, j Oyster Creek Nuclear Generating Station, December 19, 1979, by !
URS/ John A. Blume & Associates, Engineers.
j
Paga 22 The referenced analysis complies with Paragraph 3.1.2, Pages V-3-1 thru V-3-6 of the original Facility Description and Safety Analysis Report (FDSAR) (which was in force at the time of the analysis) for a Safe Shutdown Earthquake (SSE) of .22g ground acceleration. The subject paragraph requires the piping to meet an operablity criteria for SSE and the system energy absorption capacity at locations in which the operability criteria is exceeded. The referenced analysis uses operability criteria of 3S,and ductility capacities to address system operability and energy absorption capacity. These methods and criteria are widely accepted in the industry. In addition, since this analysis has been performed, the site specific ground acceleration at Oyster Creek has been determined to be .165G for an SSE , therefore, if the piping stresses of the referenced analysis are adjusted for this reduction in ground acceleration, the stresses in the piping will all be below the 3S, criteria. The updated FSAR-(circa 1984), Section 3.7.3.2, changed the seismic requirements for Class 1 piping to ANSI B31.1-1983 Edition, through Winter 1984 addenda, to address those piping systems being reanalyzed for NRC IE Bulletin 79-14/02. All other original systems not reanalyzed for 79-14/02 still meet the requirements of the original FDSAR. GPUN will amend the updated FSAR to include those systems not addressed by NRC IE Bulletin 79-14/02.
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Prg7 23 NRC CONCERN
" Deficiency D-4: (Open) Piping Specification Seismic Criteria DISCUSSION:
The team reviewed the piping specification (Reference A) which governed the design and installation of all large-and small-bore piping.outside of the Reactor Building. GPUN Budget Activity 328141P2 was part of a long-term GPUN program to monitor pipe wall thickness in ESW piping following the permanent removal of the corrosion-protective lining from portions of the piping.
Section 2.6 of the specification, Hangers, Anchors and supports, noted that:
For piping outside of containment vessel under this specification, the magnitude of the horizontal force shall be equal to 0.05 times the operating dead load of the piping. The vertical seismic load shall be considered zero.
The team noted that this seismic design criterion did not agree with the more stringent seismic requirements of FSAR Sections 3.2.1 and 3.7.2.6.
GPUN licensing did not provide the team with a formal response to this issue
- during the week of October 31, 1988. However, GPUN informally indicated that all large-bore NSR piping outside of the Reactor Building had been reanalyzed as part of GPUN's 79-14 program tg requalify NSR piping and supports to installed configurations. Therefore, the team did not consider this issue to be safety-significant for large-bore NSR piping. This program, in part, reanalyzed NSR large-bore piping and supports to the required FSAR seismic criteria. However, some NSR small-bore piping may have been installed to the less stringent seismic criterion (Reference A). In a separate finding, the team asked GPUN to confirm that NSR small-bore piping and supports were field-routed in acaordance with FSAR seismic criteria and the piping code of record for OCNGF (Reference B). This item remains open until the piping specification is corrected to reflect the criteria specified in the FSAR.
REGULATORY BASIS:
FSAR SeJtion 3.2.1 stipulates, in part, that a safe shutdown could be achieved during, a ground motion of 0.22g. FSAR Section 3.7.2.6 requires, in part, that the F.aximum vertical acceleration be taken as 2/3 of the maximum horizontal acceleration, and that the maximum horizontal and vertical accelerations be considered to occur simultaneously. Appendix B to 10 CFR 50 requires in part thtt measures be established to correctly transfer design bases to procedures, specifications, drawings and instructions. The design bases specified in the FSAR were not correctly incorporated in the GPUN piping specification for OCNGS."
G]'UN Response The paragraph referenced is from Burns & Roe Specification S-2299-48 Para.
- 2. 5-B , Page II-70. However, Addendum No. 3, Page 6 supercedes and clarifies the intent of this paragraph. The clarification shows that the seismic forces l
spetified apply only to piping which is non-seismic (seismic analysis not required). To further justify that this was the true intent of these seismic forcts and that this was consistently applied, Burns & Roe Reactor Building Piping Specification S-2299-60 Para. 2.6-B, Page II-79 clearly states these seismie forces apply only to non-Class I seismic piping.
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p; ;,. , ;'since the. seismic forces specified by-the paragraph-in. question apply,only.'to~
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g 'non-seismic. designed piping,:the paragraph is not in conflict with the FSAR i 1 sections referenced which apply.only to seismic Category 2 piping.
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1 Pign 25 NRC CONCERN
" Deficiency D-7 Containment Spray Piping Stress Analysis DISCUSSION:
GPUN Budget Activity 402876 authorized the reanalysis of NSR piping and supports in response to a 1985 NRC audit which identified documentation deficiencies in GPUN's 79-14 and 79-02 programs. These NRC programs required the qualification of piping and supports to installed configurations. The team reviewed the piping analysis for a portion of the containment spray system (Reference A). This analysis was one of 50 piping analyses that GPUN performed to requalify 11 NSR systems.
The team could net confirm that GPUN had addressed the following items in the related calculation,
- a. A comparison of the magnitudes of the seismic displacements and the as-built gaps at the locations of pipe supports 241-64 and 241-65. The piping analysis deleted these restraints due to excessive gaps.
- b. A check of the piping flange joints.
GPUN licensing did not provide'the team with a formal response to these issues during the week of October 31, 1988. However, GPUN informally indicated that the computed seismic displacement at the location of pipe support 241-65, which GPUN abandoned in place, exceeded the as-built gap dimension, and that a pipe clamp at that location will be removed to increase the gap clearance. GPUN will also document the adequacy of the pipe flanges. This item remains open.
REGULATORY BASIS:
GPUN's failure to perform documented checks of pipe support gap clearances with respect to the computed seismic displacements for supprts that were abandoned in place does not meet the intent of IE Bulletin ' Seismic Analysis for As-Built Safety-Related Piping Systems,' 79-14, which requires that piping and supports be analyzed to installed configurations. Section 110, Piping Joints, of the piping code of record for OCNGS notes, in part, that the piping joint used shall be selected with consideration of the mechanical strength."
GPUN Response (a) A review of the 691 pipe supports in the 79-14/02 scope identified 7 pipe supports which were ' abandoned-in-place'. These 7 include supports 241-64 and 241-65 which were identified in the inspection finding. The following table sumarizes the action taken on the justification for these 7 " abandoned-in-place" supports.
212-82 Rod Hanger -
See Note 212-133 Rod Hanger -
See Note
, 241-25 Rod Hanger -
See Note 241-29 Plate with Gap -
Gap Adequate 241-31 Angle with Gap -
Gap Adequate 241-64 Cantilever -
Cantilever Removed 241-65 Clamp in Frame -
Clamp Removed
.Pags 26' Notes Rod hangers are considered to be non-functional for seismic loading
.unless justified on an individual basis. This is specified in CPUN.-
Specification SP-1302-12-208, Para. 4.8, and'was acknowledged by the NRC during Inspection 85-37 as the design basis for the 79-14 reanalysis.
Therefore, seismic effects of " abandoned-in-place" rod hangers need not
.be considered.
(B) Flanged piping in the 79-14/02 scope is limited to the Emergency Service Water System piping which is a low temperature line. The flange joints satisfy the design code requirements of ANSI B31.1 1983.through Winter 1984' addenda, paragraphs 112,108.1 and 104.5.1 (A), since the flanges
.were manufactured in accordance with ANSI B16.1 and B16.5. Furthermore, since this piping was designed for all load combinations, including SSE, to ANSI B31.1 stress limits, the maximum possible result %nt bending moment is low compared to the design capacity of the fla:rge joint.
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Pcgs . 27 NRC CONCERN i
" Deficiency D-8:_ Small-Bore Pipe Support Spacing DISCUSSION:
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GPUN Budget Activity 402876 authorized the reanalysis of NSR piping and supports in response to e 1985 NRC audit which identified documentation deficiencies in GPUN's 79-14 and 79-02 programs. These NRC programs required the qualification of piping and supports to installed configurations. Section 2
. 1 of the GPUN design specification (Reference A) indicated that NSR small-bore piping (less_that 2-1/2 inches in diameter) was originally field-routed to generic spacing criteria (which GPUN cannot retrieve).
Appendix B to the specification documents GPUN's regeneration of design span lengths for small-bore piping by extrapolating the~1arge-bore piping support
. spans shown in FSAR Figure'3.7-14. GPUN compared these extrapolated span lengths with measured piping spans for the small-bore Liquid Poison System.
GPUN also performed a detailed piping analysis of the Liquid Poison System.
GPUN concluded that the original support spacing criteria used to_ install small-bore pipe at OCNGS were satisfactory and were implemented.
However, the team could not confirm that Appendix B of the specification i satsifactorily addressed the following concerns:
- a. Reduction of allowable pipe support spans for small-bore pipe due to the presence of in-line masses such as valves.
- b. Consideration of piping thermal expansion due to operating temperatures greater than 150 degrees F.
- c. Adequacy of small-bore pipe supports subjected to combined dead, thermal and seismic loads.
In a separate finding, the team noted that NSR piping outside of the Reactor Building may have been. installed to seismic criteria less stringent than current FSAR licensing commitments.
GPUN licensing did not provide the team with a formal response to this issue during the week of October 31, 1988. However, GPUN informally indicated that the detailed analysis which GPUN performed to requalify the piping and supports for the Liquid Poison System provided adequate confirmation that NSR small-bore piping and supports were properly installed at Oyster Creek. This item remains open.
The team considered this issue to be a documentation rather than a safety significant issue.
REGULATORY BASIS:
NSR Emall-bore piping and supports are required to meet the stress requirements and allowables of the piping code of record for OCNGS (Reference B)."
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l Prga 28 GPUN Response I
(a) The span review performed for the Liquid Poison System showed reasonable j seismic spans and that most in-line masses (i.e., valves) had nearby- j seismic restraints. j The adequacy of this reasonable spacing was confirmed by the dynamic computer analysis performed for this system which justified the original support spacing / location.
It was not the purpose of the span review to develop a small bore seismic spacing criteria, but rather to demonstrate that some original criteria similar to that used for the large bore tseismic piping was -
probably applied and that the application was/is justifiable by checking a sample NSR system in detail.
Therefore, the effers of in-line masses has been addressed by the fact that the reviewed system had the existing valves seismically restrained and that the restraint location was adequate based upon dynamic analysis.
All new seismic Category I small bore piping and all significant modifications to existing seismic category I small bore piping is evaluated by dynamic computer analysis and not by application of span tables.
(b) A review identified 7 safety-related small bore (<2 1/2" nominal diameter) piping lines with design temperatures greater than 150'F.
Four of these lines have beGn reanalyzed since the original design using computer analysis with satisfactory results. These four lines are:
3/4" NE-1 - Isolation Condenser Vent Piping 1 1/2" NP-2 -
Liquid Poison Pump Discharge Piping 2" NP-1 -
Liquid Poison Pump Suction Piping 2" NC-2 -
Scram Discharge Volume Piping The isometric drawings of the remaining three lines were reviewed and adequate flexibility for thermal expansion appears to be present.
1" NC-2 -
CRD Exhaust Water Piping 1 1/2" NZ-2 -
Core Spray Min. Flow Piping 2" '.tD-1 -
RV Drain to RWCU Piping (c) Burns & Roe Specification S-2299-60, Para. 2.6-B, Page II-82 states that for 2" and under piping subjected to " careful stress analysis", the same level of design applied to large bore pipe supports will be performed.
Page II-78 of the same specification identified Class I seismic systems which would have been subjected to " careful stress analysis". Included in this listing are only two small bore pipe lines both of which have been reanalyzed since the original design. These lines are in the liquid poison and CRD hydraulic systems.
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1 Therefore, while detailed original design calculations of small bore support on piping subjected to " careful stress analysis" (i.e.,
safety-related piping) are not available, it is apparent based upon the original design specification and the positive results c,f later reanalysis that there is no reasonable basis to question the adequacy of small bore safety-related pipe supports.
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!= 's PIga 30 NRC CONCERN
._" Deficiency D-9: 2-Bolt Base Plate Qualification DICUSSION :
GPUN requalified base plates.with anchor bolts under Budget Activity 328130P5 in response to a.1985.NRCfaudit which identified documentation deficiencies in LGPUN's 79-14'and.79-02 programs. .These NRC programs required the qualification of piping and supports.to installed configurations. -The team reviewed several
-pipe support configurations which consisted of pipe hangers supported by supplementary. steel. The supplementary steel was supported at one end by a.
- 2-bolt' base plate which was fixed to a wall. The base plate anchor bolts were aligned horizontally. The GPUN pipe support calculation which qualifies each of these supports modeled the connection between the supplementary steel and the base plate aus a pinned rather than a rigid _ connection. This analytical assumption provided a conservative check of the stresses in the supplementary.
ateel. However, it was unconservative when computing the loads in the base plate anchor bolts, since only the. anchor bolt shear force, but not the anchor bolt tension or: shear and tension interaction, was computed and checked against c'.lowables. ESW system pipe supports SW-2-H2, SW-2-H8, and SW-2-H10 were
' typical examples of such pipe support configurations.
GPUN did not. provide the. team with a formal response to this issue during the week of October 31, 1988.
The team considered this to be a safety-significant issue, since the NSR ESW system was required to perform a safety function during and after a SSE. This item remains open.
REGULATORY BASIS:
Anchor bolts for 2-bolt base plates subject to " weak axis" bending may not be properly qualified to.the anchor bolt acceptance criteria which GPUN specifies in. Reference a. . Appendix B to 10 CFR 50 requires design bases to be translated into specifications."
GPUN Response The 2-Bolt base plate pinned connection aerumption for shear and tension interaction has been evaluated using a finite element concept. It was concluded that the simplified analysis of a frame containing a 2-Bolt base plate (support SW-2-H2), assumed to act as a pinned connection, is correct.
Supports SW-2-H8 and H10 have been modified under the 79-02 program. They do not.have 2-bolt base plates.
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NRC CONCERN
" Deficiency E-1: . Poor Records Managemeat for Load Changes on IE Power Sources Discussion:
The team reviewed Modification Package #BA-402856. This modification package was issued to lower the reactor water level setpoint for feed water (FW) control following reactor scram. The FW pump run out protection logic design J was also changed. The purpose of lowering this setpoint was to prevent l flooding of the isolation condenser steam line after a reactor scram. The modification provided an automatic means to reduce the magnitude of reactor water level variations during plant shutdown transients not caused by a failure in the FW control system. Following a scram, voids collapse and cause a low sensed water level without actual reduction in water inventory. Without the mechanism provided by this modification, the low sensed level would provide a higher demand from the FW control system and the resultant excess inventory addition could result in high water level and possible flooding of the isolation condenser steam lines.
This modification would increase loading on the 120 volt vital power bus and on the class 1E station battery system. The team reviewed the licensee's load control program for the 120 volt vital AC bus, 125 volt DC system, and class 1E diesel generator system and noted that the licensee did not have adequate procedures to record and track the load additions. Individuals have recorded load additions simply by noting them on a scratch pad. The licensee's load control program was found to be a collection of such scratch pad papers, without signatures or proper verification, stored with individual engineers without being tracked.
The licensee failed to follow Technical Division procedures for d( gn verification which stipulated that the design verification must be performed as a control measure to verify adequacy of engineering designs which are within scope of GPUN operational QA plan 1000-PLN-7200.01. The team noted that class 1E power' sources, such as the 120 volt vital power supply, 4160 volt diesel generators and 125 volt station battery systems, were within the scope of the QA plan.
The team was informed that the licensee was aware of this concern and was in the process of installing a computer-based tracking system. In the future, this system would have the capacity to track all load additions to all affected buses. The tracking system would also indicate the remaining margin of the available capacity of the associatrJ equipment. The licensee expects to complete this activity in the nere future. Until this PC-computer-based tracking system is installed, the licensee committed to rearrange all the note pad papers showing loading chang 0s and to make these documents official by signing, verifying, and controlling Umm. This item remains open.
REGULATORY BASIS:
ANSI N-45.2.11, Section 6.1, states, 'The results of design verification efforts shall be clearly documented with the identification of verifier clearly ,
indicated thereon, and filed. Documentation of the results shall be auditable against the verification methods indicated by the design organization.'
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Pcg3 32 Section 6.2, Design Verification, states that, where changes to previously verified designs have been made, design verification shall be required for the changes, including evaluation of effects of these changes on the overall )
design."
GPUN Response GPUN has been keeping records for load changes to all electrical plant busses.
Exhibit 4.2 of Technical Functions Procedure EMP 014 provides an Electrical Loading Data form, which is completed by the project sponsor. The Electrical Power Manager approves or rejects the new loads based upon the magnitude of the load and available capacity on the proposed bus. The applicable calculation is revised periodically. The revised calculation follows applicable procedures including Design Verification Procedure EP-009. Therefore, GPUN disagrees with the NRC claim that the load control program is not being tracked or verified.
We believe that good engineering practice is followed and the applicable calculation is reviewed periodically.
GPUN will ensure that load calculations are updated prior to equipment installation to ensure that thuse loads are acceptable. Load records are not kept informally on scratch pads. Electrical engineers maintain individual notebooks containing up-to-date load information. This approach will be enhanced by utilizing a PC-based tracking system which will provide uniformity in data presentation. Additional procedural guidance will be developed by 12/31/89 to insure that a uniform approach is used in load control.
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.: n Paga 33' NRC CONCERN
, " Deficiency E-2: Failure to Calculate Acceleration Time of the Pump and Motor Assembly for the Core Spray Pump and Emergency Service Water Pump.
DISCUSSION:
The team reviewed mini-modification package No. 7. The mini-mod packages were
' issued by the branch of the design engineering located at the plant. The purpose of this modification was to replace the thermal overload and instantaneous relays device numbers 49 and 50 respectively, with new relays having more reliable and predictable operating characteristics, along with a shorter reset time.
The team noted that the trip settings of the new relays were established using calculation No. C-1302-750-5350-001, dated February 10, 1988. The team reviewed this calculation and noted that the settings were justified without accounting for the acceleration time for the pump and motor assembly during startup. In a situation where these motors were fed from the diesel generator, it might be possible that, due to block start of ESF loads and random loads, the minimum voltage at the motor terminals may be less than the normal rated voltage. On low terminal voltage, motors take a longer time to accelerate, and during acceleration motors draw currents higher than normal running currents.
Therefore, if the acceleration time of the pump and motor assembly is more than the trip time of the over-current relay, the relay will trip before the motor attains its full speed.
Licensee engineers informed the team that, since information re2ated to the motor torque, load torque, and moment of inertia of pump-motor assembly was not retrievable, it was not possible to compute acceleration timing. The team ;
expressed a concern that without knowing acceleration time, it is not possible l to verify that these pumps will not be tripped prematurely during acceleration. This created an indeterminate situation in which it is possible that core spray and essential service pumps disabled by premature trip may not be available to mitigate demands of a concurrent LOCA and LOOP incident.
On the last day of the inspection, the licensee revised the calculation and demonstrated that the selected settings were adequate for a terminal voltage equal to 80 percent of the motor-rated voltage. The calculation inputs were based on test rest 1ts which in the team's opinion may not be conservative.
Testing was done only by establishing recirculation flow, and full flow was not established. The team concluded that for full flow, the acceleration time will be greater than the value used in the revised calculation. The calculation also assumed that the_ load torque remains constant throughout the acceleration period, which may not be true. The team believes that the load torque value will increase during the early phase of the acceleration period, making net available torque lower than the assumed value in the calculation. Lower values of net torque may result in increased time for acceleration. The calculation also assumes a terminal voltage equal to 80 percent of the motor-rated voltage. This basis has not been substantiated. The testing results did not indicate the accuracy with which acceleration time was measured and does not indicate the range of speeds within which between this time was noted. The team considered this item open.
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P g3 34 REGULATORY BASIS:
Unavailability of an ESF system for mitigation of a concurrent LOCA and LOOP event.is contrary to the Technical Specification requirements which stipulate ]
that these pumps must be available under all plant conditions other than cold l l
shutdown."
l GPUN Response.
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- 1. Starting times and appropriate motor / pump data are not available for these motors.
- 2. When the normal power supply is used, these pumps do not have a starting ;
problem due to undervoltage conditions because on-line automatic voltage I regulators have been installed. 1
- 3. When starting these pumps from the emergency diesel generators, GPUN does not expect unacceptably long acceleration times. During the inspection, GPUN presented a methodology for calculating motor acceleration times. Following discussion of this approach, calculation
- C-1302-700-5350-003 was revised as discussed. With this, we believed this issue to have been resolved.
Our response to the specific points raised by the above concern is as follows:
In the revised calculation, (Rev. 1), two equations were generated for pump starting time. These were derived from the generic equation for starting time
't's t = Wh2x rpm 308 x T Reference 11-40 (1), Large Apparatus Induction Motors Switchgear & Control Handbook - R. W. Smeaton This equation assumes that the net average torque T (motor torque - pump torque) remains constant throughout the acceleration period.
We agree that this may not be true in the case of an actual pump start. In ,
fact, motor torque varies from ' acceleration' torque to ' pull' torque and !
steadily ascends to the higher ' break' torque and finally at rated speed levels off to a much lower value of ' full' load torque.
In relation to the pump, torque decreases in the early acceleration period (first 10% of the acceleration period) and then steadily increases to full load torque.
The torque value inferred in our calculation is net starting torque which, in fact, may be lower than the net average torque over the whole acceleration range. The starting time thus calculated should be longer than actual and our results are conservative.
From time-current curves TCC-M1A & M2A, it is quite clear that there is approximately 100% safety margin between the calculated acceleration time (approximately 6 seconds) and the '51' overcurrent relay trip times (approximately 12 seconds) at the starting current value. This margin
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p Ppga 35' compensates for.the approximations used in the calculations of the starting times. These approximations include potential undervoltage (less than 80%) and pump' flow conditions.
It should also be noted that the calculations include motor thermal data, and
'the overcurrent relays are set just below the thermal damage curves. This setting allows maximum motor acceleration times and still protects the motor from damage.
Based on this, we believe that in lieu of actual data for these pumps and motors,'the methodology of Rev. 1 of our calculation presents a reasonable approach to. calculate starting time. In conclusion, we believe that there is adequate assurance that the pumps will accelerate without tripping overcurrent relay 1 during start.
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- P,ig3 36 NRC CONCERN
" Deficiency E-3: Potential for the Process Variable of the Safety Systems to Exceed the Technical Specification Limit Without Being Noticed by the Plant operator.
DISUSSION:
The team reviewed three sample calculations for process trip setpoint. These calculations were done for modification packages No. BA-402896, BA-402256 and BA-408761 and were performed by Burns & Roe, Ebasco Services and the in-house staff, respectively.
Calculation No. 4283-12-11-001 (modification package BA-402896) for the reactor
. pressure trip setpoint showed a margin of 10 psig between the nominal setpoint and the Technical Specification limit. This margin consisted of total. loop
- . uncertainty of 8,psig plus an additional allowance of 2 psig. The margin was divided into three zones., Zone 1, known as the 'AS LEFT' zone, was between the nominal setpoint and a margin of.3 psig above. During calibration, the setpoint could be left up to the upper limit of this zone. Zone 2 began at the upper limit of Zone 1 and extended to the 'as-found' allowable limit'of the setpoint. This upper limit was 2 psig below the technical specification limit.
-In a situation where a setpoint has been left on the upper limit of the 'AS LEFT' zone (i.e., greater than 3 psig above the nominal setpoint), it may be possible for the process variable to exceed the Technical Specification limit although by virtue of loop inaccuracy, the measured value may still appear to be within the 'as-found' allowable limit.
Licensee engineers were made aware of this situation. During follow-up.
discussions with the licensee engineers, the team was informed that a similar methodology has been used for calculation and calibration of other safety-related instrument loops. Consequently, the team was concerned that this may be a generic problem.
In the latter part of the inspection, the team was informed by the licensee that the in-house review committee had already identified this concern. As a
. result, the licensee intended to revisit all potentially affected safety-related instrument loops before the restart of the plant. On the last day of this inspection GPUN informed the team that there are eight such loops which may require revised calculation. This item remains open.
REGULATORY BASIS:
The Technical Specification specifies maximum and minimum (as applicable) values of the process variables of safety systems. To ensure the safety of the plant and the public, these processes are required to operate within established limits'during normal plant operation."
GPUN Response
' Calculation No. 4283-12-11-001 (Modification Package BA-402896) for the reactor pressure trip setpoint was performed by Burns and Roe. The calculation was l done in accordance with GPUN Engineering Standard ES-002. The standard did not l
require inclusion of calibration tolerance while determining the set points.
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'*'.u p .- -Pega.37 The referenced calculation was revised to reduce the' calibration tolerance from 3 psi.to 2 psi. This would ensure that the technical specification limit fer reactor high pressure will not be exceeded with the calculated drift. The
. procedures were developed using the revised calculation.
'The results of the review of the eight calculations are summarized in the attachment to GPUN Letter 5000-89-1704 to NRC dated January 19, 1989. GPUN Engineering Standard ES-002, " Instrument Error Calculation and Setpoint ,
Determination", has been revised to' include calibration tolerance while .j determining setpoints.
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P2g2 38 j HFC CONCERN
" Deficiency E-5: Failure to Consider Random Loads on the Diesel Generator During LOCA and LOOP Sequences.
DISCUSSION:
New therr.al and instantaneous over current (device number 49 and 50) relays were installed per mini-modification No. 7. For setpiont calculation of these relays, diesel generator test data was used to establish minimum available voltage during worst-case diesel loading. The team reviewed the test data and emergency diesel generator'- .equenced loading for LOCA and LOOP events.
Through this review, the team noted that, during the sequencing of required loads, it is possible that random loads may be started along with the sequenced loads. Random loads are those loads which are started automatically by the process signals and are not bypassed by the accident signal. Examples included sump level activated pumps, and thermostatically controlled heat tracing. As a result, the OCNGS generator may be loaded to approximately 10 percent of its rated capacity. The team noted that the licensee failed to perform an analysis of the effects of random load initiation on the diesel generator's voltage and frequency drop and its recovery between loading steps.
The licensee informed the team that the possibility of starting of such random loads along with the sequenced loads exists, but that the size of the individual loads was small. The licensee also contended that randomly started loads would not create an appreciably harmful effect on the voltage and frequency of the diesel generator. The licensee further informed the team that GPUN intended to analyze the condition in the near future. This item remains open pending completion and results of the GPU analysis.
REGULATORY BASIS:
ANSI-N45.2.11 states that procedures shall include requirements for documenting assumptions and identifying assmptions that must be verified as the design proceeds. GPUN made a tacit (undocumented) assumption that random loads would not affect the voltage and frequency of the diesel generator. Random loads, if ,
not bypassed by the ' sequencer active' signal, may try to start along with a '
sequenced step load and may trip the diesel generator or impede its voltage and frequency dip within the assigned time. This, in turn, could delay or prevent starting of the subsequent ESF load."
GPUN Response A load analysis will be completed by December 31, 1989, to determine the effect of starting of the random leads during worst case diesel generator (DG) loading sequence. It is expected that the starting of the random loads will not adversely affect the operation of the DG.
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.r Pago 39' j I
1 NRC CONCERN
" Deficiency E-4: Calculation Errors, Such as Unverified Assumptions and Failure to Account for Response Time of the System Equipment.
DISCUSSION:
1 The team reviewed three sample calculations for process trip setpoint. These )
calculations were done for modification packages BA-402896, BA-402256 and BA-408761 and were performed by Burns & Roe, Ebasco Services, and the in-house I
staff.respectively. As a result of this review, the team identified the following concerns.
- a. The safety margin value used in the reactor pressure setpoint calculation lacked justification.
- b. Instrument accuracy, as stated by the manufacturer, was assumed to be representative to three standard deviations.
- c. Power supply voltage variation was assumed to be within +5 percent without proper references.
- d. Instrument field calibration accuracy was assumed to be +0.25 percent of span and setpoint without verification from the plant site.
- e. Calculations did not account for the response time of the instrument loop or the equipment which is actuated by the accident signal. During discussion with the licensee's engineers, the team was informed that General Electric may have considered this factor in their calculation for the Technical Specification limit of the process. Therefore, as long as the original values of the Technical Specification limits are used, it may not be necessary to reevaluate response time. The team expressed concern that a modification to the system may change its response time and, therefore, the effects of the changed response time on the trip settings for modified systems should be evaluated.
The licensee's engineers informed the team that, in the future they will exercise additional caution in verifying the calculations performed by the architect-engineer companies. The team also noted that the licensee lacked a program to track and resolve unverified assumptions in calculations performed for safety-related systems.
The team was informed by licensee that the in-house review committee had already identified the concern related to failure to consider the response times and intended to resolve this concern by revisiting all safety-related setpoint calculations before the restart of the plant. This item remains open.
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Paga 40 I
REGULATOR'l BASIS: ,
ANSI-N45.11 requires that procedures shall include requirements for documenting j assumptions and identifying those assumptions that must be verified as the design proceeds. Failure to resolve unverified assumptions in the safety-related calculations and to consider the effects of equipment response time changes due to modification may result in incorrect settings for activation of the safety systems. These incorrect settings may compromise safe l shutdown capability of the safety system and may eventually compromise public j safety." ]
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GPUN Response
- a. The intent of the calculation was to verify that the technical specification limit will not be exceeded with the existing setpoint and calculated drift. The safety margin value derived in the calculation was the margin between the calculated setpoint (accounting for drift) and the technical specification limit. In this calculation, the instrument setpoint for the analog loops was assumed to be the same as the setpoint for the switches that the analog system replaced.
Therefore, a basis for this safety margin value was neither warranted nor identified.
- b. GPUN Engineering Standard ES-002, " Instrument Error Calculation and Setpoint Determination", Section 6.1, provides guidance to assume vendor published accuracies to be 3 sigma confidence level. The basis for this guidance is in accordance with attachment 'A' to NRC memorandum from Mr. Peter S. Kapo, DSI to Mr. Walter R. Butler, Chief Containment System Dranch, DSI dated August 23, 1982.
- c. GPUN procedure EP-006, " Calculation", Section 4.1.4, requires
" assumptions and basic data shall be identified with justification, as appropriate."
The instrument cabinet power supply that provides power to all the components of the instrument loop is fed by the main generator during normal power operation. oyster Creek operating procedure 336.1 in step 3.3.1.2. requires that the main generator voltage be maintained within 25.5KV and 22.8KV (i.e. +6.25 and -5.00 percent of 24 KV nominal). The power supplies for the electronics cabinets have voltage regulation.
Therefore, the actual output of the cabinet power supplies is maintained within the specified 5 percent and the assumption is conservative.
- d. GPUN Procedure EP-006, " Calculation" Section 4.1.4, requires
" assumptions and basic data shall be identified with justification, as appropriate." The calculation reviewed by the SSOMI team has been revised to reference the calibration procedure es the basis for the calibration accuracy number used in the calculation.
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- e. The modifications performed are judged not to significantly alter response time. The issue of response time t esting was reviewed as part of the Systematic Evaluation Program (SEP) under SEP Topic VI-lO.A, Testing of Reactor Trip System and Engineered Safety Features, Including-Response-Time Testing. The NRC Staff conclusions regarding this topic a:e presented'in the Integrated Plant Safety Assessment Report (NUREG 0822, January 1983), Section 4.26.1. The conclusion reached was that functional testing, as performed at Oyster Creek, is sufficient to demonstrate function and that response time testing is not required.
Functional teste continue to show acceptable performance for these instrument loops. The reactor high pressure setpoint calculation has been revised to include the basis for not computing response time.
The comment that GPUN lacks "...a program to track and resolve unverified assumptions in calculations performed for safety-related systems" is addressed as follows:
GPUN Procedure EP-009-Design Verification, Section 4.3.4, provides guidance to verification engineers to identify concerns which require further investigation. Section 5.1.4 requires that the Responsible Section Manager (who has the responsibility to approve design verification) prepare Task Requests for items requiring further investigation. We consider this adequate to track and resolve concerns.
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