ML17332A970
| ML17332A970 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 10/06/1995 |
| From: | John Hickman NRC (Affiliation Not Assigned) |
| To: | Fitzpatrick E INDIANA MICHIGAN POWER CO. |
| References | |
| TAC-M89878, TAC-M89879, NUDOCS 9510170050 | |
| Download: ML17332A970 (28) | |
Text
o (htober 6, 1995 Hr.
E.
E. Fitzpatrick, ice President Indiana Michigan Power Company c/o American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, OH 43215
SUBJECT:
DONALD C.
COOK NUCLEAR PLANT, REQUEST FOR ADDITIONAL INFORMATION CONCERNING PROPOSED EMERGENCY ACTION LEVELS (TAC NO. H89878 AND M89879)
Dear Hr. Fitzpatrick:
We have completed our review of your March 31,
- 1995, response to our November 22, 1994, request for additional information (RAI).
Several issues remain to be resolved before we can complete our review of the D.C.
Cook emergency action level methodology.
We request that you provide additional information as discussed in the enclosed RAI to support resolution of the remaining issues.
To help ensure that you clearly understand the issues raised in the RAI and to support prompt resolution of your request, I have suggested to your staff that we have a conference call to clarify the RAI and subsequently have a meeting to discuss your responses.
If you have any questions on this issue, please contact me at (301) 415-3017.
This requirement affects nine or fewer respondents and, therefore, is not subject to the Office of Management and Budget review under P.L.96-511.
Sincerely, Origiral signed by Docket Nos.
50-315 and 50-316 cc w/encl:
See next page John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation DISTRIBUTION:
'ocket Fil~ek PUBLIC PDI II-I r/f JRoe EAdensam (e)
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(A)D:PD31 NAME CJamerson DATE 10/I/'p /95 JHickm n:db 10/ 6 /95 BHol i an 10/
/95 95iOi70050 9510ih PDR ADQCK 05000315 PDR 2 Boj. 69 "FcICIAL R CORD COPy
Mr. E.
E. Fitzpatrick Indiana Michigan Power Company CC:
Regional Administrator, Region III U.S. Nuclear Regulatory Commission 801 Warrenville Road Lisle, Illinois 60532-4351 Attorney General Department of Attorney General 525 West Ottawa Street
- Lansing, Michigan 48913 Township Supervisor Lake Township Hall P.O.
Box 818 Bridgman, Michigan 49106 Al Blind, Plant Manager Donald C.
Cook Nuclear Plant 1 Cook Place Bridgman, Michigan 49106 U.S. Nuclear Regulatory Commission Resident Inspector's Office 7700 Red Arrow Highway Stevensville, Michigan 49127 Gerald Charnoff, Esquire
- Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.
W.
Washington, DC 20037 Mayor, City of Bridgman Post Office Box 366 Bridgman, Michigan 49106 Special Assistant to the Governor Room 1
State Capitol
- Lansing, Michigan 48909 Nuclear Facilities and Environmental Monitoring Section Office Division of Radiological Health Department of Public Health 3423 N.
Logan Street P. 0.
Box 30195 Lansing, Michigan 48909 Donald C.
Cook Nuclear Plant Mr. S. Brewer American Electric Power Service Corporation 1 Riverside Plaza
- Columbus, Ohio 43215
UPPLEIVIENTAL RE UEST DDITIONALINF RIVIATI N E A DIN D.:
K
&2M LE LAN EAL REVI I
NS T NUIVIAR NE P-7 ETH DOL Y
The NRC completed a review of the additional information transmitted in the March 31, 1995, D.C. Cook Nuclear Plant emergency action level (EAL) submittal.
The additional information was requested after the initial NRC review of the D. C.
Cook Nuclear Plant June 13, 1994, EAL submittal.
The proposed EALs and additional information submitted were reviewed against the guidance in NUMARC/NESP-007, "Methodology for Development of Emergency Action Levels," Revision 2.
NUMARC/NESP-007 has been endorsed by the NRC in Regulatory Guide 1.101, "Emergency Planning and Preparedness for Nuclear Power Reactors," Revision 3, as an alternative means by which licensees can meet the requirements in 10 CFR 50.47(b)(4) and Appendix E to 10 CFR Part 50.
Since the staff has previously endorsed the guidance in NUMARC/NESP-007, the review focused on those EALs that deviated from the guidance.
The additional information submitted and the procedure/plan review resolved many of the staff's questions.
However, several areas of concerns remain with specific EALs which require further information and review.
Please provide this additional information as discussed below. The comments are organized with the original NRC comment first, the D. C. Cook response, and then the NRC response.
NRC G n r
I Commen:
The D. C. Cook Nuclear Plant EAL scheme deviated from the NESP methodology by not grouping EALs under initiating conditions.
The D. C. Cook Nuclear Plant EAL scheme included two separate tables; one table contained all the initiating conditions and a second table contained all the EALs. In most cases the EAI s were exact duplicates of the initiating conditions.
The separation of EALs from initiating conditions was confusing and would make classification more difficult.
D.
r n
The initial submittal was not clear concerning the purpose of the Initiating Condition (IC) table.
The table was developed to replace the previous information in the Emergency Plan Section 12.3.5, Emergency Classification System.
The table would have the appropriate level of detail for off-site agencies.
The use of the IC table in the Emergency Plan simplifies the process for revising EALs by not requiring a change to the Emergency Plan for minor changes that do not change the relevant IC. The response also stated that the term IC and EAL are not used in the revised EPIP table.
Within the table, each Emergency Condition Category is displayed with its associated Recognition Category on one page to maintain the ability to see the emergency classification escalation pathway.
NR Ev I
in fD..
kR 1.
The staff accepts D.C. Cook's proposal to include only the ICs in the Emergency Plan and include the detailed EALs in a Emergency Plan Implementing Procedure (EPIP). However, the focus of the NRC comment was not on whether the detailed EALs should be included in the Emergency Plan but rather that the ICs should be included as a part of the EAL scheme in the EPIP.
As stated in the NUMARC guidance, an EAL is "a pre-determined, site-specific, observable threshold for a plant Initiating Condition that places the plant in a given emergency class."
The NUMARC scheme groups EALs under the IC to which the EALs correspond.
This allows the person classifying (and the people being notified of the classification) to understand the plant condition of concern.
This logical grouping of EALs under ICs was not included in the D.C. Cook EAL scheme.
In addition, the D.C. Cook EAL scheme in some cases included NUMARC ICs as EALs. Also, the revised IC table included in the D.C. Cook Emergency Plan contains some EALs in addition to ICs.
The D.C. Cook EAL scheme contained in the EPIP should be revised to group EALs under ICs.
The EAL scheme must be described in the Emergency Plan but need not be included in the plan.
The IC table is acceptable as a description of the EAL scheme but should be revised to be consistent with the ICs contained in the EAL scheme.
Please provide information regarding actions taken to address these concerns.
2.
In addition to the concerns stated above, the staff is concerned that the "and/or" logic specified by NUMARC has not been used.
Numerous EALs lack specific "and/or" wording prescribed by NUMARC which would force the declaration for a given EAL condition. Wording used by D. C. Cook includes the phrases symptoms include" and "such as" which does not clearly indicate classification is required when one or more of the conditions is present.
Further, it is not clear whether one of the listed conditions, all of the conditions, or a combination of the conditions is needed for classification.
The D. C. Cook Nuclear Plant Deviation Basis Document (DBD) does not contain justification for the deviation.
I Provide further justification for these deviations from the NUMARC/NESP-007 guidance.
~RI:
IC <
15 mrem/hr on R-5 as an indicator for an Unusual Event (UE) declaration. This indicator'has been moved to ECC R-3 because it is more indicative of loss of water level. Emergency classification escalation will occur via ECC R-3 if one or more irradiated fuel assemblies is uncovered or if R-5 exceeds 1000 mrem/hr. NRCEval i n fD, kR The D. C. Cook response is not acceptable without additional information or clarification. The fact that plant design does not provide for remote monitoring of spent fuel pool or refueling canal level is not sufficient justification for eliminating AU2, EAL P2, regarding an uncontrolled water level decrease in the spent fuel pool and fuel transfer canal. Indications such as operator observations of a specific amount of level decrease should be included as an EAL for this IC. Provide further justification for these deviations from the NUMARC/NESP-007 guidance. ~MR ~ A: (Corresponding to D. C. Cook EAL R-1 and NUMARC EAL AA1, "Unplanned Release of. ~.Radioactivity" ) D. C. Cook Nuclear Plant EAL did not include the Note" specified in the NUMARC EAL. The assessment specified in this EAL is to ensure that the release exceeds 200 times the technical specification value based on the actual amount of radioactive effluent being released. D. C. C k r n The D. C. Cook response is the same as the response to NRC Comment 1A above. NRCEv lu i n fD. n The response is not acceptable without further information or clarification. As discussed under NRC comment 1A above, the note appears to be applicable to D. C. Cook. In addition, in this revision, D. C. Cook has added an additional EAL for the boundary dose exceeding 10 mrem/hr total effective dose equivalent (TEDE) without explanation. Lacking an updated DBD for the current submittal, it was not clear why this EAL was added and how the TEDE rate is to be calculated. IC 0 EI ~
- 0. 0'. C k EAL II-I 2 NUMAIICEAI.
AA1, "Unplanned Release of...Radioactivity" ) As previously discussed under NUMARC AU1, EALs 1 and 2, D. C. Cook Nuclear Plant eliminated the usage of sample analysis (AA1, EAL ¹2) as an EAL without proper justification. n The D. C. Cook response was the same as for NRC Comment 1B above. R Evl in fD. kR n The D. C. Cook response is not acceptable as explained under NRC Comment 1B above. Please provide further justification or clarification for deviations from NUMARC/NESP-007 guidance. ~NC: IC p dI g
The D. C. Cook Nuclear Plant DBD stated, "Generic EAL ¹1 is not used since there are not area monitors in all areas where fuel uncovery can occur." Provide details of the arrangement of area radiation monitors in the vicinity of fuel handling and storage areas and provide additional justification for not using these radiation monitors in an EAL under this IC. D.,C kr n The details of the radiation monitoring system can be found in the UFSAR, Chapter 11; the D. C. Cook Nuclear Plant does not have a refuel floor area radiation monitor with remote readings. There is an area monitor (R-5) in the Fuel Handling Building and general area monitors in upper and lower containment. However, the key phrase in the generic IC is loss of water level." If a valid reading for the containment general area radiation monitors were included in the EAL scheme, confirmation of the loss of water level would still need to be specified. Since reactor cavity.water level at the Cook Nuclear Plant is checked locally, the statement addressing decreasing water level is the best available classification guide. R Ev I i n fD kR n The D. C. Cook response is not acceptable. The lack of remote reading area radiation monitor does not provide justification for eliminating NUMARC AA2, EAL ¹1. Radiation monitors readings in or near fuel handling areas (with local or remote 0 ~ readings) should be included as site-specific EALs corresponding to NUMARC example EAL 0'1. The NUMARC basis states that each site should also define its EALs by the specific area where irradiated fuel is located such as Reactor Cavity, Reactor Vessel, or Spent Fuel Pool. It was not clear how this was done for the D. C. Cook EALs. The redundancy in NUMARC examples EALs may be beneficial in preventing missed classifications. NUMARC AA2 examples are purposely redundant with an assigned or" logic (1 or 2 or 3 or 4). Provide additional justification for deviations from NUMARC/NESP-007 guidance. ~NR ~ 4 (Corresponding to D. C. Cook EAL R-2 and NUMARC EAL AA2, "Major Damage to Irradiated Fuel...") As in prior D. C. Cook Nuclear Plant EALs associated with reactor cavity and spent fuel pool water level, R-2 deviates from the NUMARC guidance by not including instrumentation or visual level limits. Provide additional justification for this deviation from the NUMARC/NESP-007 guidance. D. C. C k r n The visual verification of water level is possible through observation of level markings on the side of the, spend fuel pool and reactor cavity. NR Eval a ion f D.C. C ok R n The D. C. Cook response did not address the original question of why the EALs associated with reactor cavity and spent fuel pool water level deviate from NUMARC guidance by not including 'ns rum n ati n r vi I I vel limi . The response did not provide justification for the deviation. If visual verification of water level is possible (as stated in the response) and level markings are on the side of the spent fuel pool and reactor cavity (as stated in the response), it appears that NUMARC guidance could easily be implemented. Provide justification for the deviation from NUMARC/NESP-007 guidance. 4: lC <<4 dl g O. C. C 4 EAL ll-2 4 ~ CNARC EAL AA2, "Major Damage to Irradiated Fuel...") D. C. Cook Nuclear Plant EALR-3 combines a requirement (not suggested in the NUMARC generic EAL) that there be visual indication of damage to irradiated fuel" combined, using an AND statement, with high alarms on any of several radiation monitors (which were claimed to be lacking as a justification for eliminating NUMARC generic EAL AA2-1). While the visual indication" aspect of this EAL is satisfactory, and implied in the NUMARC generic IC, if not in the generic EALs, combining the "visual indication" with the alarms is non-conservative. Ifthis EAL were split into two pieces, it would meet the basic intent of NUMARC generic EALs AA2-1 and AA2-2. D.. kr n indications from area radiation monitor R-5 were moved into this ECC because valid increasing radiation levels on this monitor indicate loss of water covering spent fuel. In addition, suitable indicators under the SAE column for ECC R-3 were added that address indications of loss of water level that can uncover multiple spent fuel assemblies or indication of spent fuel in the spent fuel area losing its water cover based on readings from monitor R-5. This provides a clear escalation path from Unusual Event (uncontrolled loss of water level) to Alert (single spent fuel assembly can become uncovered) to SAE (multiple fuel assemblies can become uncovered). The new ECC is simpler and more in-line with the intent of the NUMARC guidance. ECC R-4 was changed to state in the Alert column: Anyindication ofmajor damage to irradiated fuel such as: ~ Unexpected increase in airborne radioactivity during fuel handling operations. NR Ev I i n fD.. kR n The D. C. Cook methodology for this EAL is acceptable. However, no information on plant specific monitors, methods, and trigger levels which make this EAL readily observable and measurable were included in this EAL. In addition, justification for the EAL setpoint for the SAE classification for damage to multiple fuel assemblies, i.e. 1000 mrem/hr, was not provided. Provide additional information for deviations from NUMARC/NESP-007 guidance. ~NRC EA: lg 6 dl 6 C. C. C 6 ECC 11-1 d MCMARE EAL AS1, "Boundary Dose...exceeds 100 mRem...") The D.C. Cook EAL scheme did not include an EAL corresponding to EAL AS1-1. The justification given was that dose assessment could not be performed within the first 15 minutes of the release. Licensees are expected to have the capability to quickly perform dose assessments. D kr n The radiation monitor EAL corresponding the NUMARC EAL AS1-1 was modified to include the following: For an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> event, a valid reading > 7.5 E-2 uclcc on VRS-1507/2507 may exceed this threshold. Evl in D kR The basis for the use of an 8-hour event (or 8-hour default release duration) is not clear or explained in the D. C. Cook DBD. The NUMARC guidance recommends that 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> be assumed when calculating the radiation monitor setpoints. In addition, it is not clea'r why the radiation monitor EAL setpoint is not a factor of 10 less than the setpoint for the General Emergency EAL. Furthermore, no discussion was included regarding the appropriateness of this setpoint as it relates the setpoint for the Alert radiation monitor EAL. As discussed under NRC Comment 1A, a note needs to be included in this EAL to ensure that dose assessment results will be used to classify the event if the dose assessment results are available and to ensure that, if assessment activities extend beyond 15 minutes, the event will be classified based upon radiation monitor readings. Provide the rationale for the use of an 8-hour event duration and justification for values used in the radiation monitor EAL. ~NR: IC <<5 dI II O. C. C 3 ECC 5 5 d NCMARC EAL EIIR. "Loss of Most or All Safety System Annunciators") The D. C. Cook Nuclear Plant IC Deviation stated, Deleted reference to loss of 'most'nnunciators and loss of all or most indicators. The loss of 'most'nnunciatorsis not plausible with our design. The only single failure that affects a majority of the annunciators willcause the loss of 'ALL'nnunciators. Loss of indications was deleted from this IC because a maj or loss ofindications (CRIDs) willalso render the compensatory non-alarming indications inoperable (SPDS, PPC) because loss of the CRIDs willdisable input to these systems also. The D. C. Cook Nuclear Plant DBD did not provide a full discussion of the power supplies and interrelationships of control indicators (CRIDs), SPDS, PPC and annunciators. However, based on the DBD discussion, it would appear that the units could suffer a loss of most or all indication (CRIDs) and therefore SPDS and PPC, without a loss of annunciators. While the NUMARC example EAL starts with a statement on loss of annunciators, the BASIS discussion clearly states that loss of either annunciators or indications are to be considered. A more thorough discussion of the possible causes of loss of various sections of the safety system sections of the annunciator tiles is required in order to evaluate the licensee's claim that a loss of "most" annunciators is not plausible. Provide additional information on annunciator and indicator design, and justification for the apparent deviation from NUIVlARC/NESP-007 guidance. D kr n D. C. Cook now uses the NUMARC standard nomenclature of most." In addition, annunciator panels associated with safety systems are listed in a footnote. With the exception of the theoretically possible, but practically impossible, discrete failure of each indication downstream of its isolation amplifier (involving dozens of unrelated failures), it is not possible for either Cook Nuclear Plant unit to lose indications without also losing the associated annunciators because of the bistable interface between indication and alarm. The design of the GRID system is described in the UFSAR. However, based on a re-review of the proposed EALs, Cook Nuclear Plant determined that its classification scheme is more complicated than necessary. Accordingly, ECC S-7, Loss of Alarms or Indication, has been revised to state: Unusual Event Unplanned loss ofmost safety system annunciatorsin a unit for ) 15 minutes. ~ Panels 104-114, 119, 120 ~ Panels 204-214, 219, 220 Alert Loss of one or more CRIDS resulting in a significant plant transient. -OR= Unplanned loss of most safety system annunciatorsin an unit for > 15 minutes with a significant plant transient. Site Area Emergency Loss ofALL CRIDs. NR Eval ai n fDC. o kR on The staff accepts that it may be difficultto lose indications without losing annunciators if the interface bistable, amplifier, or isolation device taps off downstream, and is located in the same vicinity as the indications. However, the-NUMARC basis states that "itis further recognized that most plant designs provide redundant safety system indication powered from separate uninteruptable power supplies. While failure of a large portion ofannunciatorsis more likely than a failure of a large portion ofindications, the concern is includedin this EAL due to difficultyassociated with assessment ofplant conditions." The intent is to have this EAL recognize the potential (however slight) for a loss of indications. The revised EAL for the UNUSUAL EVENT did not incorporate any reference to a loss of indications and is therefore slightly more conservative than the NUMARC EAL. The ALERT EAL addresses a loss of one or more CRIDs "resulting in a significant plant transient." NUMARC guidance references a loss "during a plant transient." While it may be likely that the loss could cause a plant transient (and thus an Alert classification would be appropriate), the transient does not need to be caused by that loss for classification as the D. C. Cook EAL states. The D. C. 10 r Cook EAL is more limiting (less conservative) with regard to transients because it does not envelop plant transients that may have been in progress coincident with the loss of CRIDs. Please provide further information or clarification for deviations from NUMARC/NESP-007 guidance. R mm n 1 A: (Corresponding to D. C. Cook ECC S-3 and NUMARC EAL SU7, "Loss of Required DC") The NUMARC BASIS discussion for site specific" loss states, "Bus voltage should be based on the minimum bus voltage necessary for operation of safety related equipment. This voltage value should incorporate a margin of at least 15 minutes of operation before the onset of inability to operate those loads." D. C. Cook Nuclear Plant added a 15-minute time requirement to its definition of loss of 250V DC buses. No explanation was provided as to how the 210-volt figure was computed. The intent of the NUMARC EAL is to initiate an emergency declaration at the voltage for which 15 minutes of load capacity remains, not arrive at that voltage and then wait 15 minutes before making a declaration. D kr n As documentedin a Cook Nuclear Plant system description (SD-DCC-PS104), the 210V DC limit was based on battery service test acceptance criteria. The 15 minute margin was appliedin the wrong direction and has been eliminated. Based on interpolation, the voltage at 15 minutesis about 213V DC. However, 213V DC is not readable on installed control room instrumentation due to meter scaling (10V DC divisions on a dialindicatorJ. We have decided to use a value of 215V DC becauseit is the closest value that can be read on the control board. Use of the 215V DC valueis furtherjustified because slowly decreasing battery voltage should only occur during loss of all ACpower scenarios, which are separately classified by ECC S-2A and S-28. Other events of concern for this ECC would involve bus loss ii.e., 0 V DC bus voltage) and are readily discernible against the threshold value of 215V DC." NR Ev I aion. fD .Co kR n The response for this comment is acceptable except that the D. C. Cook response states that a value of 215V DC will be used because it is the closest value that can be read. However, procedure PMP 2080.EPP.101, Emergency Classification, Revision 1, Attachment A, page 2 of 5, note 2 states Loss of DC power occurs whenever a DC train's voltage is less than 210V DC." The EAL value (210V DC), according to the D. C. Cook response, is not readable in the control room. Please provide clarification for the apparent inconsistency on the use of 210V DC footnoted to the EAL matrix. 11 ~NR (3 3 3 3 C.C.C RECCE-31 R Ev I i n f D. n The technical information given in the D. C. Cook response for this comment is acceptable; however, it appears that the EAL Unplanned loss of SFP cooling for greater that 60 minutes in ALLmodes" upon which the NRC comment was based has now been removed from the Emergency Plan and PMP 2080.EPP.101, Revision 1 without explanation in the D. C. Cook response. Provide confirmation on whether or not the EAL will be included in the D. C. Cook EAL scheme. R mmn 11: (Corresponding to D. C. Cook ECC S-5 and NUMARC EAL SA4, "Loss of...Annunciators") Previously in the D. C. Cook Nuclear Plant DBD (under NUMARC SU3) the licensee .claimed that, "Loss of indications was deleted from this IC because a major loss of indication (CRIDS) will also render the compensatory non-alarming indications inoperable (SPDS/PPC) because loss of CRIDs will disable input to these systems also." The NUMARC EAL, stated simply, would be met by one of two basic conditions: (1) Loss of annunciators AND.loss of control indications, SPDS, PPC, etc., or (2) Loss of annunciators AND a transient in progress. The D. C. Cook Nuclear Plant EAL would require a loss of annunciators, control room indication, and a transient in progress. This D. C. Cook Nuclear Plant EAL, contrary to licensee statement under NUMARC SU3, seems to imply the SPDS and PPC can continue to operate with a failure of CRIDS. As written, the D. C. Cook Nuclear Plant EAL is either non-conservative (if SPDS and PPC can continue to operate with a loss of CRIDs), or contains a logic error (if SPDS and PPC are rendered inoperable by CRIDs failure). Provide additional information on the relationship of CRIDs, SPDS, and PPC and justification of the apparent deviation from the NUMARC/NESP guidance. D.. ookr s n e: The D. C. Cook response referenced responses to NRC Comment 9 above and NRC Comment 12 below. NR Evl in fD.. o kR n The referenced responses did not contain complete explanations which would answer the inconsistencies discussed in NRC comment 11. The effort to understand the licensee response and this EAL (and others) was hampered somewhat by the lack of a revised D. C. Cook Emergency Plan Classification vs 12 NUMARC/NESP-007 Deviation Basis Document which reflected information on the changes made to the EALs since the original submission. Provide additional information on all annunciator EALs and justification for any deviations from the NUMARC/NESP guidance. Include information on the power supplies, failure modes for control room indications, annunciators, and compensatory non-alarming indications (SPDS, PPC, etc.). R mmn 2: (Corresponding to D. C. Cook ECC S-3 and NUMARC EAL SS3, "Loss of AllVital DC Power" ) D. C. Cook Nuclear Plant has reduced the classification level of this NUMARC example IC/EAL to an Alert. The Alert level EAL, S-3 has been discussed previously, as it relates to operation in Modes 5 and 6, under NUMARC generic EAL SU7. In the D. C. Cook Nuclear Plant IC Deviation statement, the only apparent difference between Cook and other PWRs, might be the fact that D. C. Cook Nuclear Plant has inverters for the normal power supply to control room instrumentation. As in the evaluation of the D. C. Cook Nuclear Plant response to NUMARC SU7, additional information on DC power supplies is required to complete this evaluation. The deviation statement should be written in terms of what sets D. C. Cook Nuclear Plant apart from other Westinghouse PWRs of the same vintage with respect to this NUMARC IC/EAL. D.. ook res n The D. C. Cook response and DBD indicated that control room indications would not be lost as normal power to instruments is from inverters fed from AC power. Containment cooling, which would isolate on a loss of DC power, could be manually unisolated to restore cooling. The DBD stated that the loss of DC power would also cause a complete failure of automatic actuation for all safeguards equipment and loss of control room annunciators and if in Mode 1, a rapid RCS cooldown due to failure of the automatic turbine trip systems. Ev I in fD.. kR n The D. C. Cook response for this comment is not acceptable without further information. The NUMARC basis for this EAL includes both the ability to mgri~ir and u)~nrol plant safety functions. The loss of DC power at D. C. Cook appears to degrade some aspects of both control and monitoring and manual compensatory actions would be required to regain control. Provide additional information for deviations from NUMARC/NESP-007 guidance. 13 (Corresponding to D. C. Cook ECC R-1 and NUMARC EAL SS6, "Loss of...Annunciators") For the reasons previously discussed for the Unusual Event and Alert EALs for loss of annunciators, this EAL does not appear to be appropriate criteria for indication of the inability to monitor a significant transient.. D.. kr n The D. C. Cook response references the response to NRC Comment 9 above. R Ev I in fD.. kR n The response will require additional information as indicated in NRC Comments 9 and 11 above. NR mmn 1 (Corresponding to D. C. Cook ECC R-1 and NUMARC EAL HA1, "Turbine Failure...") The D. C. Cook Nuclear Plant EAL deviates from the NUMARC guidance by not specifying turbine missiles and by not including the condition of missile penetration. Provide justification for this deviation. D, kr n The conditions in the Alert column for ECC N-7, Equipment or Structural Failure, have been revised to state: ~ Turbine-failure missile affects the operability ofgghhr systems required for the current operating mode gr for safe shutdown. ~ Significant visible damage to plant structures due to any cause. The new wording is consistent with the Alert level indicators proposed for ECCs '-1 through N-6 in the response to NRC Comment 18 above. To clarify the escalation from the Unusual Event threshold (penetration of the affected turbine casing), the indicator states either some other system must be affected or that significant structural damage outside the affected component must occur. For more information concerning required equipment, please refer to the response to NRC Comment 20 below. NR Ev I i n fD.. kRes n The D. C. Cook response is not acceptable. NUMARC guidance for Alert, HA1, example EAL 06 states: 14 Turbine failure generated missiles resultin any visible structural damage to or penetration of any of the followingplant areas: (site-specific) list. and the NUMARC basis states in part: "ifmissiles have damaged or penetrated areas containing safety-related equipment the potential exists for substantial degradation of the level of safety of the plant." D. C. Cook's EAL is less conservative and deviates from NUMARC guidance. The NUMARC intent is to declare the Alert if there is penetration into a vital area of the plant, not withstanding damage to systems/components within the area. The D. C. Cook EAL would require turbine missiles to affect system operability or cause significant visible damage to the specified areas before an Alert would be declared. I Provide additional information for deviations from NUMARC/NESP-007 guidance. ~NRC C 2: IC <<2 dt 2 O..C. C 2 ECC It-4 4 NIINARC EAI HS2, "Control Room Evaluation" ) The NUMARC BASIS for this EAL states, "In cold shutdown and refueling modes, operator concern is directed toward maintaining core cooling such asis discussed in Generic Letter 88-17, Loss of Decay Heat Removal." In power operation, hot standby, and hot shutdown modes, operator concern is primarily directed toward maintaining critical safety functions and thereby assuring fission product barrier integrity." The D. C. Cook Nuclear Plant EAL limits control outside the control room to RCS inventory Control," thus ignoring the other critical safety functions.'rovide justification for this deviation from the NUMARC/NESP-007 guidance. D.. C k res ons: The EAL has been revised to state: Control Room evacuation has occurred ~NO control of the followingprocesses is not established within 30 minutes: ~ Reactivity ~ RCS inventory ~ RCS temperature ~ SG heat sink and a footnote was added to explain the meaning of the term control." NR Evl in fD kR n The D. C. Cook revised EAL and response for this comment are not acceptable. The time period for establishing control was lengthened from 15 minutes to 30 15 minutes in the revised EAL without justification or an explanation of the technical basis. The NUMARC basis clearly states this time should not exceed 15 minutes. This EAL would be acceptable if the time to establish control was changed back to the original 15 minutes. Revise the EAL to meet NUMARC guidance or provide additional detailed justification for the deviations from NUMARC/NESP-007 guidance. ~A ~ 2: IC<<F dfg A F<<FAIL fg IgldgALI The NUMARC criterion for Fuel Clad Barrier Example EAL 0 1 states: POTENTIAL: Core Cooling - ORANGE QB Heat Sink - RED The D.C. Cook equivalent EAL stated: POTENTIAL: Core Cooling - ORANGE Heat Sink - REO - AND - Wide range levelin at least 3 SGsis (29% (43% for adverse containment) The D.C. Cook DBD provided discussion on how the SGs will continue to act as a heat sink until wide range level is lost. Although absent from the DBD, the D.C. Cook FPB matrix does contain an OR statement for the two CSFs. The generic Westinghouse EPGs for the HEAT SINK CSF require no SGs in the narrow range, and feedwater flow less than the safeguards AFW flow requirement for heat removal to equate to a RED path. The D.C. Cook DBD did not make a statement as to whether it had deviated from the Westinghouse EPGs in formulating its plant-specific EOPs (i.e., do the EOPs contain this same caveat on SG wide-range level in order to achieve a RED path on the HEAT SlNK CSFP). Provide justification for this deviation from the NUMARC/NESP-007 guidance in the form of discussion on criteria for the plant-specific HEAT SINK RED path on the CSF Status Trees (CSFSTs), and by providing details on the relationship of wide to narrow range SG level instrumentation. D.. kr n Cook Nuclear Plant does not deviate from Westinghouse Owners Group (WOG) Emergency Response Guidelines (ERGsJ for the heat sink critical safety function status tree (CSFSTJ. Narrow range SG level covers 144 inches down from the same upper tap as the wide range SG level. Wide range SG level Clcovers about 25 feet from SG tube sheet to the bottom of the dryer assembly in the top of the SG. Both ranges of SG level are environmental qualified for a post-LOCA containment. The SG narrow range levelis calibrated for normal operating 16 pressure and temperature conditions. The SG wide range levelis calibrated for cold shutdown conditions. The Cook Nuclear Plant specific symptoms are written in such a way that they are not misinterpreted or misapplied to result in declaration of a site area emergency (due to potential loss ofboth fuel clad and RCS barriers) following a normal reactor, trip. The WOG ERGs for the Heat Sink CSF require no SG levelin the narrow range and feedwater flowless than the safeguards auxiliary feedwater (AFWJ flow requirement for heat removal to equate to a RED path. Following a reactor trip from fullpower, the normal transient response for Cook Nuclear Plant is for SG level to drop below the bottom of the narrow range. In addition, main feedwateris automatically shut offand the AFWpumps willstart on a low level signal.
- Thus, due to automatic response to a normal reactor trip, a transitory set of conditions which mimic the Heat Sink CSF - RED conditionin the period prior to startup of the AFWpumps can exist.
According to the 'Background for WOG ERG F-O, 'age 4, the Heat Sink CSF protects both the Fuel Clad barrier and the RCS barrier. However, Review of WOG ERG Background for FR-H. 1, Section 2, shows that Heat Sink CSF - RED does not immediately represent a severe challenge to these two, as shown in the description of the loss of feedwater accident with no operator action. Review ofFigure 1 and the description of Period 3 of this event makes it clear that RCS heat-up to pressurizer (PZRJ PORV operation does not occur for between 25 to 50 minutes following the loss of feedwater event. Core uncovery therefore cannot result solely due to loss ofsecondary heat sink until the PZR PORV has dumped a large quantity ofreactor coolant without adequate RCS makeup. The WOG ERG Background for FR-H. 1 further states on page 11 that '(a)ll cases show that core uncoveryis minimized and long term core cooling is sustained through RCS bleed and feed heat removal. 'n page 4 1, it further states that 'symptoms of SG dryout provide adequateindication to successfullyinitiate feed and bleed. 'herefore, symptoms that requireinitiation of feed and bleed cooling constitute a 'potential loss'hallenge to-core cooling and thus the Fuel Clad barrier. Cook Nuclear Plant EOP FR-H. 1 uses 3 out of4 SG WR levels less than 29% (43% for adverse containment, conditions) as the bleed and feed initiation criterion. Inclusion of this requirementin addition to a Heat Sink CSFST - RED condition therefore: (1J appropriately focuses the concern on the ability to maintain longer term core cooling, and (2J assures that the Heat Sink CSFST-RED willnot be misapplied and resultin unnecessary declaration of a site area emergency. NRCEv l ai n fD,C. kRe on The D. C. Cook response is not acceptable. The staff understands the technical discussion and justification, but does not consider the nonconservative deviation from NUMARC guidance necessary. With minor wording changes, which do not deviate from the intent of NUMARC, the EALs can be tailored for a specific plant 17 response. To suggest that a licensed shift supervisor might declare a CSFST RED, then an SAE during a reactor trip with normal plant response due to the wording of this EAL is not reasonable. The momentary loss of SG levels (below narrow-range indication) and loss of feed (before AFW pumps start) is a very short transient condition. If D. C. Cook believes that there is a clear chance for misinterpretation and declaration of an SAE for a reactor trip with normal plant response, then words that exclude the momentary transient could be added without resorting to the less conservative AND"condition (3 out of 4 wide range steam generator level, etc.). Provide additional justification for deviation from NUMARC/NESP-007 guidance. R mm n 2 A: (Corresponding to Fuel Clad Loss EAL) The D.C. Cook DBD provided the following EAL Deviation statement: We have applied a time limiton the use of containment radiation monitors because the radiation levels for a given amount of fuel damage willbe constantly decreasing with time. In'order to prevent the SS/SEC from having to refer to a time-plot of radiation vs fuel damage, we have limited the viabilityof this symptom and used . the lowest radiation reading within that time frame associated with the level of core damage assumedin the generic guidance. Therefore this symptom is not used on the Fission Product Barrier table associated with ICs. D..C okrs on The context of the basis statement in question is that a constant amount of core damage willresult in a continuously decreasing containment radiation monitor reading due to radioactive decay. The original statement can be misinterpreted, and did account forincreasing core damage. The basis statement was revised to state: 'A containment radiation monitor value corresponding to a noble gas release to containment from 5% clad damage decayed for 90 minutes was selected for thisindicator of fuel clad loss. This radiation monitor reading corresponds to an undecayed noble gas release due to about 2% to 3% clad damage. Thus, the containment monitor value corresponds to the 2% to 5% clad damage range described by the generic guidance. 'rom an organizational standpoint, immediate core damage is the worst case condition because assessment personnel are not going to be available for up to 60 minutes. Ninety minutes was selected because it allows adequate time for the Technical Support Center to be staffed and core damage assessment to begin. While the 200 chris a valid assessment ofimmediately occurring core damagein the range of 2% - 5%, the 90 minute threshold does not need to be included. Therefore, theindicator was revised to state: CNTMTarea radiation ) 200 Ruhr prior to the TSC beginning core damage assessment. 18 NRC Evaluation of D.C. Cook Response: The D. C. Cook response for this is acceptable ifthe EAL "TSC assessment of core damage >5% clad failure" is modified to "assessment of core damage >5% clad failure" and note 2 at the bottom of the EAL table "prior to the TSC beginning core damage assessmerit" is deleted. The control room shift technical advisor may perform a core damage assessment prior to the TSC being activated so it is not appropriate to limit the EAL for dose assessment to those performed by the TSC only. Furthermore, the EAL "CNTMTarea radiation >200 R/hr" should apply at all times because when the containment radiation levels are above this value core damage is indicated. Provide additional justification for the deviations from the NUMARC/NESP-007 guidance. NR mm n 2A; (Corresponding to RCS Barrier Loss EAL) On the loss EAL, we added the caveat that subcooling cannot be restored. This allows some leeway in those circumstances where the EOPs direct minimizing subcooling for some mitigating action. The licensee did not provide an example of where the EOPs purposefully have the operators reduce subcooling to (30 degrees F (it is presumed that this includes instrument error). The licensee placed no time constraint on the phrase "can NOT be restored." D.. Cook r ons: The D. C. Cook response stated: The Cook Nuclear Plant EOPs, which are based on the WOG ERGs, have Sl re-initiation criteria based on loss ofsubcooling. At that point, operator action would be taken to restore subcooling. There are instancesin which the EOPsinstruct the operators to purposefully lower RCS pressure to reduce subcooling to less than minimum required. For example, ECA-3., directs the operators to purposefully reduce RCS pressure to saturated conditions. Since subcooling lossis due to operator action, then the generic EAL indicator ofloss of subcooling due to RCS leak rate exceeding available RCS makeup would not apply for this case. Including a time limitcouldimply that action should be undertaken to restore subcooling, which is not appropriate when EOP ECA-3.2 isin effect. Itis noted that our previous submittal made use of the requirement that RCS subcooling remain <30 degrees fahrenheit. That number referred to the sum of pressure and temperature measurement system errors, including allowances for normal channel accuracies, translated into temperature using saturation tables. Upon close examination, we do not believe this single number to be reflective of our current procedures which specify different values depending on the system conditions and the severity of the environmentin the containment. Since several 19 different numbers areinvolved, andin fact can change as we change plant instruments, we do not believe it to be appropriate to include a dynamic tabulation of such values as part of the emergency planning documentation. We have, therefore, elected to change the subcooling criteria for the RCS barrier to be 'less than the minimum required. 'his minimum requiredis intended to refer to the value currently specific by the Cook Nuclear Plant emergency operating procedure which is applicable at the time. Itis also noted that this changein wording willmaintain emergency planning documentation on thisissue consistent with the recommendations of the WOG ERGs which recommend that the minimum subcooling be set equal to instrumentinaccuracy. We also believe that is important to retain the phraseology 'and cannot be restored'ince its removal could result in unnecessary agitation of operations under emergency conditions. In addition to the example cited above where current procedures may require the plant to deliberately be taken out of a required subcooling condition, itis also possible to envision conditions where subcooling may be inadvertently lost, but can be restored at willsince the RCS barrier remains intact. Itis not prudent to unnecessarily perturb emergency response operations under either of these conditions. Evl in fD. The D. C Cook EAL and response are not acceptable without additional information. Although the D. C. Cook logic is understandable, the deviation from NUMARC guidance is not necessary or appropriate. It is recognized that during accident mitigation the plant must be operated in accordance with approved EOPs. NUMARC recognized this when formulating initiating conditions and example EALs. Tp NNMARC pl ICIEAL << "~RC L k R GREATER THAN II Pl makeup capacity as indicated by a loss of RCB subcooling. D. C. Cook did not d "~LkR GREATER THAN k p p I y' I EAL. I I C. C. Cook EAL, the loss of subcooling statement stands alone without being connected to the RCS leak rate. For the NUMARC EAL, if the loss of subcooling is deliberate, in accordance with procedures, and there is no RCB leak rate greater than makeup capacity, classification would not be required (based on subcooling atone). At that point in the EOPs (intentional reduction of subcooling), the SEC/SS should be I RC dl I . Tp f, P "~RGB L k R GREATER THAN k p capacity" (which D. C. Cook omitted from the EAL) would not be present and no classification should be required. If the NUMARC EAL was implemented without omitting part of the NUMARC wording, it would appear the NUMARC EAL would be appropriate for D. C. Cook as written. Provide additional information for the deviation from NUMARC/NESP-007 guidance. 20 BB: IC <<2 BI II RCBB <<I 2 <30 minutes, 'as based on the assumption that the steam flow was coming from a relief valve. However, itis clear that thisintent could have been 24 misinterpreted using the original wording for thisindicator. Therefore, Containment Loss indicator was revised to state: Any RUPTURED SG has unisolable steam flow out of the containment. By using the term 'unisolable,'the revised EAL would resultin an immediate determination of containment barrier 'loss'or breaks outside containment and stuck open atmospheric relief valves that cannot be isolated, rather than waiting for 30 minutes, as would be possible using the original wording. Further, the SG PORVs are i'solable. 'hus, this wording excludes normal and expected response of the ruptured SG PORV during the initialRCS depressurization phase of EOP E-3, which was the originalintent using the 30 minute threshold, and the intent of the generic guidance. R Evlu in fD, k The D. C. Cook response and revised EAL are unacceptable. The criterion for the NUMARC EAL is primary-to-secondary leakage greater than technical specification allowable. D.C. Cook did not provide justification that a leakage rate of greater than technical specification corresponds to a ruptured steam generator tube as defined the EOPs. Furthermore, the D. C. Cook EAL specifies that the release to the environment is "unisolable," whereas the NUMARC EAL specifies simply that there is a release. The D. C. Cook response did not adequately justify this deviation. Provide justification, for this deviation from the NUMARC/NESP-007 guidance. ~C3 (Corresponding to Containment Barrier Potential Loss EAL) The D. C. Cook Nuclear Plant DBD did not adequately justify the equivalence of the D. C. Cook Nuclear Plant EAL and the NUMARC EAL. In particular, no information was given regarding the relationship between the NUMARC criteria of "core exit thermocouples in excess of 700 degrees F and reactor vessel level below top of active fuel" and the D. C. Cook Nuclear Plant EAL, Core Cooling CSFST - RED." D.. kr n The Cook Nuclear Plant Core Cooling CSFST (based on the WOG ERG CSFST for plants with RVLIS)includes two separate paths that resultin Core Cooling - RED. The firstis core exit temperature greater than 1200 degrees F. This is used by all Westinghouse plants, whether or not they have a RVLIS design like that at the Cook Nuclear Plant. The second RED path forplants with a RVLIS design such as that at the Cook Nuclear Plant is core exit temperature greater than 700 degrees F and RVLIS Level less than 3.5 ft from the bottom of the active fuel. These are the two conditions referred to in the generic methodology. 25 However, either core cooling RED path willresult in entryinto EOP FR-C. 1,
Response
to Inadequate Core Cooling, which is the applicable functional restoration procedure referred to in the generic methodology.
Thus, use of Core Cooling CSFST - RED is directly equivalent to the indicators usedin the generic methodology.
R Ev I
i n f D.
k R n
The D, C. Cook response for the original comment is acceptable.
However, the D. C. Cook EAL contains a wording deviation which is more conservative than NUMARC. The NUMARC EAL, Containment Barrier, Core Exit Thermocouple Readings, under potential loss states in part:
"restoration procedures not effective within 15 minutes."
The D. C Cook equivalent EAL states in part:
"core temperature does not drop within 15 minutes."
The NUMARC basis (pg. 5-33) states the functional restoration procedures are those emergency operating procedures that address the recovery of the core cooling critical safety functions.
It further states the procedure is considered effective if the temperature is decreasing (as the D. C. Cook EAL states) or if the vessel water level is increasing.
Although the D. C. Cook EAL is more specific in stating what the NUMARC restoration procedure being "effective" means, the NUMARC basis information regarding a vessel level increase was not included.
Therefore NUMARC's "effective" means temperature decrease or level increase and D. C. Cook's "effective" means only a temperature decrease.
Provide additional information as to whether the omission of the "vessel level
'ncrease" for this EAL was intentional and if so, the technical basis for the omission.
26